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abstract | An attenuator system for attenuating a radiation beam, including a first attenuating element placed in a path of a radiation beam for attenuation thereof, a second attenuating element placed distal to the first attenuating element for further attenuation of the radiation beam, a first positioner operatively connected to the first attenuating element, which moves the first attenuating element along a first direction, a first processor operatively connected to the first positioner for controlling motion of the first attenuating element, a second positioner operatively connected to the second attenuating element, which moves the second attenuating element along a second direction, and a second processor operatively connected to the second positioner for controlling motion of the second attenuating element, wherein a two-dimensional attenuation distribution of the first attenuating element varies linearly with respect to at least one coordinate. |
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description | The present invention relates to a MgF2—CaF2 binary system sintered body for a radiation moderator and a method for producing the same, and more particularly, to a MgF2—CaF2 binary system sintered body for a radiation moderator having a compact structure suitable for a moderator to restrict the radiation velocity and energy of radioactive rays of every kind such as neutrons and a method for producing the same. Among fluorides, a calcium fluoride (CaF2) single crystal body, a magnesium fluoride (MgF2) single crystal body and the like have been used in the optical field, for example, in the vacuum ultraviolet region of wavelengths of 160 nm or less, or in the extreme infrared region of wavelengths of 3 μm or more. They have been used as a lens, a prism and the like for such specific wavelength regions wherein light does not pass through the glass members widely used in the market such as a very-high-purity quart glass and optical glasses. Therefore, they are naturally expensive optical members. Generally speaking, there are very few cases where a fluoride is used for other than such optical uses. The CaF2 single crystal body, a lithium fluoride (LiF) single crystal body, or an aluminum fluoride (AlF3) single crystal body has been rarely used as a shield to neutrons, one of radioactive rays. However, such single crystal bodies have plane orientation dependency of moderation performance originated in crystal orientation and ununiformity due to structural defects such as subgrains, and moreover, they are extremely expensive. The radioactive rays are roughly classified into alpha (α)-rays, beta (β)-rays, gamma (γ)-rays, X-rays, and neutrons. The power passing through a substance (penetrability) gradually increases in this order. The neutrons which have the highest penetrability among them are further classified into the below-described groups, for example, according to the energy level which they have. The energy each type of neutrons has is shown in parentheses, and the larger the value is, the higher the penetrability is. In the order of the lowest penetrability, they are classified into cold neutrons (up to 0.002 eV), thermal neutrons (up to 0.025 eV), epithermal neutrons (up to 1 eV), slow neutrons (0.03-100 eV), intermediate neutrons (0.1-500 keV) and fast neutrons (500 keV or more). Here, there are various views concerning the classification of neutrons, and the energy values in the parentheses are not precise. For example, there is a view that mentions 40 KeV or less, which is within the above energy region of intermediate neutrons, as the energy of epithermal neutrons. The typical effective utilization of neutrons is an application to the medical care field. In particular, the radiation therapy in which tumor cells such as malignant cancers are irradiated with neutrons so as to be broken has been coming into general use in recent years. In order to obtain medical effects in the present radiation therapy, neutrons of a certain high energy must be used, so that in the irradiation of neutrons, the influence on a healthy part other than an affected part of a patient cannot be avoided, leading to side effects. Therefore, in the present situation, the application of the radiation therapy is limited to severe patients. When a normal cell is exposed to high-energy neutrons, its DNA is damaged, leading to side effects such as dermatitis, anemia due to radiation and leukopenia. Furthermore, in some cases, a late injury may be caused some time after treatment, and a tumor may be formed and bleed in the rectum or the urinary bladder. In recent years, in order not to cause such side effects and late injuries, methods of pinpoint irradiation on a tumor have been studied. Examples thereof are: “Intensity Modulated Radiation Therapy (IMRT)” in which a tumor portion is three-dimensionally irradiated accurately with a high radiation dose; “Motion Tracking Radiation Therapy” in which radiation is emitted to motions in the body of a patient such as breathing or heartbeat; and “Particle Beam Radiation Therapy” in which a baryon beam or a proton beam each having a high remedial value is intensively emitted. The half-life of a neutron is short, about 15 min. The neutron decays in a short period of time, releases electrons and neutrinos, and turns into protons. And the neutron has no charge, and therefore, it is easily absorbed when it collides with a nucleus. The absorption of neutrons in such a manner is called neutron capture, and one example of an application of neutrons to the medical care field by use of this feature is the below-described “Boron Neutron Capture Therapy (hereinafter, referred to as BNCT)”, a new cancer therapy which is recently gaining attention. In this BNCT, by causing tumor cells such as malignant cancers to react with a boron drug which is injected into the body by an injection, a reaction product of a boron compound is formed in the tumor portion. The reaction product is then irradiated with neutrons of an energy level which has less influences on a healthy part of the body (desirably comprising mainly epithermal neutrons, and low-energy-level neutrons being lower than epithermal neutrons). And a nuclear reaction with the boron compound is caused only within a very small range, resulting in making only the tumor cells extinct. Originally, cancer cells easily take boron into them in the process of vigorously increasing, and in the BNCT, by use of this feature, only the tumor portion is effectively broken. This method was proposed about 60 years ago. Because of small influences on a healthy part of a patient, it has been attracting attention as an excellent radiation therapy since quite long before and has been researched and developed in varied countries. However, there are wide-ranging important problems on the development such as development of a neutron generator and a device for a selection of the types of neutrons to be remedially effective, and avoidance of influences on a healthy part other than an affected part of a patient (that is, formation of a boron compound only in a tumor portion). Therefore, the method has not come into wide use as a general therapy. Significant factors in terms of apparatus why it has not come into wide use are insufficient downsizing of the apparatus and insufficient enhancement of its performance. For example, there is a latest system of the BNCT, which a group with Kyoto University as the central figure has been promoting (Non-Patent Document 1 and Non-Patent Document 2). This system comprises an apparatus for medical use only, having a cyclotron accelerator as a neutron generator which is exclusively installed without being attached to an existing nuclear reactor. One report says that the accelerator alone weighs about 60 tons, and its size is quite large. In the cyclotron system, protons are accelerated by use of a centrifugal force in a circular portion of the cyclotron and caused to collide with a target metal such as a plate made of beryllium (Be) so as to generate fast neutrons. In order to efficiently generate neutrons, it is required to make the diameter of the circular portion large so as to obtain a large centrifugal force. That is one of the reasons why the apparatus is large. Furthermore, in order to safely and effectively utilize the generated radiation (mainly neutrons), a radiation shield such as a shielding plate (hereinafter, referred to as a moderator) is required. As moderators, polyethylene containing CaF2 or LiF, as well as Pb, Fe, Al and polyethylene, are selected. It cannot be said that the moderation performance of these moderators is sufficient, and in order to conduct required moderation, the moderator becomes quite thick. Therefore, the moderation system device portion including the moderator is also one of the reasons why the apparatus is large. In order to allow this BNCT to come into wide use in general hospitals, hereinafter, downsizing of the apparatus is necessary. In addition to further downsizing of the accelerator, to improve the remedial values by developing a moderator having high moderation performance and achieve downsizing of the moderation system device by the improvement of moderation performance is an urgent necessity. The moderator which is important for downsizing a BNCT apparatus and improving remedial values is described below. As described above, in order to safely and effectively utilize radiation, it is necessary to arrange a moderator having the right performance in the right place. In order to effectively utilize neutrons having the highest penetrability among radioactive rays, it is important to accurately know the moderation performance of every kind of substances to neutrons so as to conduct effective moderation. One example of the selection of particle beam types in order to effectively utilize neutrons for medical care is shown below. By removing high-energy neutrons which adversely influence the body (such as fast neutrons and a high-energy part of intermediate neutrons) as much as possible, and by further reducing extremely-low-energy neutrons having little medical effect (such as thermal neutrons and cold neutrons), the ratio of neutrons having high medical effects (such as a low-energy part of intermediate neutrons and epithermal neutrons) is increased. As a result, a particle beam effectively utilized for medical treatment can be obtained. The low-energy part of intermediate neutrons and epithermal neutrons have a relatively high invasive depth to the internal tissues of a patient. Therefore, for example, in the case of irradiating the head with the low-energy part of intermediate neutrons and epithermal neutrons, without craniotomy required, as far as the tumor is not present in a considerably deep part, it is possible to carry out effective irradiation to an affected part in an unopened state of the head. On the other hand, when the extremely-low-energy neutrons such as thermal neutrons are used in an operation, because of their low invasive depth, craniotomy is required, resulting in a significant burden on the patient. In order to improve remedial values in the BNCT, it is required to irradiate an affected part with a large quantity of neutrons comprising mainly epithermal neutrons and some thermal neutrons. Specifically, an estimated dose of epithermal neutrons and thermal neutrons required in cases where the irradiation time is in the order of one hour, is about 1×109 [n/cm2/sec]. In order to secure the dose, it is said that as the energy of an outgoing beam from an accelerator being a source of neutrons, about 5 MeV-10 MeV is required when beryllium (Be) is used as a target for the formation of neutrons. The selection of particle beam types through moderators of every kind in a neutron radiation field for BNCT using an accelerator is described below. A beam emitted from the accelerator collides with a target (Be, in this case), and by nuclear reaction, high-energy neutrons (fast neutrons) are mainly generated. As a method for moderating the fast neutrons, using lead (Pb) and iron (Fe) each having a large inelastic scattering cross section, the neutrons are moderated to some extent. In order to further moderate the neutrons moderated to some extent (approximately, up to 1 MeV), optimization of the moderator according to the neutron energy required in the radiation field is required. As a moderator, aluminum oxide (Al2O3), aluminum fluoride (AlF3), calcium fluoride (CaF2), graphite, heavy water (D2O) or the like is generally used. By injecting the neutrons moderated nearly to 1 MeV into these moderators, they are moderated to the epithermal neutron region of the energy suitable for BNCT (4 keV-40 keV). In the case of the above Non-Patent Document 1 and Non-Patent Document 2, as moderators, Pb, Fe, polyethylene, Al, CaF2 and polyethylene containing LiF are used. The polyethylene and polyethylene containing LiF among them are used as moderators for safety (mainly for shielding) which cover the outside portion of the apparatus in order to prevent leakage of high-energy neutrons out of the radiation field. It can be said that it is appropriate to moderate the high-energy part of neutrons to some extent using Pb and Fe among these moderators (the first half of the stage of moderation), but it cannot be said that the second half of the stage of moderation using Al and CaF2 after the moderation to some extent is appropriate. That is because the moderators used in the second half of the stage thereof has insufficient shielding performance to fast neutrons, and a high ratio of fast neutrons having a possibility of bad influences on healthy tissues of a patient is left in the moderated neutron type. By reason of CaF2 having insufficient shielding performance to the high-energy part of neutrons as a moderator used in the second half of the stage thereof, part of them passes without being shielded. The polyethylene containing LiF used with CaF2 in the second half of the stage thereof covers over the entire surface except an outlet of neutrons on the treatment room side. It is arranged so as to prevent whole-body exposure of a patient to the fast neutrons, without having a function as a moderator on the outlet of neutrons. For information, the polyethylene among the moderators in the first half of the stage thereof covers over the entire surface of the periphery of the apparatus except the treatment room side, like the polyethylene containing LiF in the second half of the stage thereof, and it is arranged so as to prevent the fast neutrons from leaking to the surroundings of the apparatus. Therefore, instead of CaF2 as a shielding member to fast neutrons in the second half of the stage thereof, the development of a moderator which can shield and moderate high-energy neutrons while suppressing the attenuation of intermediate-level-energy neutrons required for treatment has been desired. From various kinds of researches/studies, the present inventors found a MgF2 sintered body or MgF2 system substances, more specifically, a MgF2—CaF2 binary system sintered body as a moderator which makes it possible to obtain neutrons (neutrons of the energy of 4 keV-40 keV) mainly comprising epithermal neutrons in anticipation of the highest remedial value, from the above neutrons moderated to some extent (the energy thereof is approximately up to 1 MeV). As the MgF2 system substances, a MgF2—LiF binary system sintered body, a MgF2—CaF2—LiF ternary system sintered body other than the MgF2—CaF2 binary system sintered body can be exemplified. As of now, there has been no report that magnesium fluoride (MgF2) was used as a moderator to neutrons, not to mention that there has been no report that a MgF2 sintered body or a MgF2—CaF2 binary system sintered body was used as such neutron moderator. The present inventors has filed an application of an invention relating to a sintered body of MgF2 simple (a technical term related to raw material technology, a synonym for “single”) prior to this invention (Patent Document 1: Japanese Patent Application No. 2013-142704, hereinafter, referred to as the prior application). MgF2 is a colorless crystal, having a melting point of 1248° C., a boiling point of 2260° C., a density (i.e. true density) of 3.15 g/cm3, a cubic system and a rutile structure according to a science and chemistry dictionary. On the other hand, CaF2 is a colorless crystal, having a melting point of 1418° C., a boiling point of 2500° C., a density (i.e. true density) of 3.18 g/cm3, a Moh's hardness of 4, a cubic system and a fluorite structure. A single crystal body of MgF2 has high transparency, and since high light transmittance is obtained within a wide range of wavelengths of 0.2 μm-7 μm and it has a wide band gap and high laser resistance, it has been mainly used as a window material for eximer laser. Or when a MgF2 single crystal body is deposited on the surface of a lens, it shows effects of protection of the inner parts thereof or prevention of irregular reflection. In either case, it is used for optical use. On the other hand, since the MgF2 sintered body has low transparency because of its polycrystalline structure, it is never used for optical use. Since the MgF2 sintered body has high resistance to fluorine gas and inert gas plasma, a few patent applications concerning an application thereof to a plasma-resistant member in the semiconductor producing process have been filed. However, there is no publication or report that it was actually used in the semiconductor producing process. That is because the MgF2 single crystal body has a strong image of extremely high price and the MgF2 sintered body produced by a general method has low mechanical strength as described in the below-mentioned Patent Document 2. As for a MgF2 sintered body, according to the Japanese Patent Application Laid-Open Publication No. 2000-302553 (the below-mentioned Patent Document 2), the greatest defect of ceramic sintered bodies of fluoride such as MgF2, CaF2, YF3 and LiF is low mechanical strength. And in order to solve this problem, the invention was achieved, wherein a sintered body compounded by mixing these fluorides with alumina (Al2O3) at a predetermined ratio can keep excellent corrosion resistance of the fluorides as well as obtain high mechanical strength. However, as for the corrosion resistance and mechanical strength of the sintered bodies produced by this method, in any combination, the sintered bodies are simply allowed to have just an intermediate characteristic between the characteristic of any of the fluorides and that of alumina. No sintered body having a characteristic exceeding one's characteristic superior to the other's has been obtained by compounding. In addition, their use is limited to high corrosion resistance uses, greatly different from the uses of the present invention. Another sintered body mainly comprising MgF2 is described in Japanese Patent Application Laid-Open Publication No. 2000-86344 (the below-mentioned Patent Document 3), but its use is also limited to a plasma-resistant member. In the Patent Document 3, a sintered body comprises a fluoride of at least one kind of alkaline earth metals selected from the group of Mg, Ca, Sr and Ba, in which the total amount of metallic elements other than the alkaline earth metals is 100 ppm or less on a metal basis, the mean particle diameter of crystal grains of the fluoride is 30 μm or less, and the relative density is 95% or more. However, the materials in the list (Table 1) of Examples of the Patent Document 3 were obtained by firing a fluoride of each single kind of the above four alkaline earth metals (i.e. MgF2, CaF2, SrF2 and BaF2), and no fired mixture of those fluorides is described. Still another example of an application of a sintered body mainly comprising MgF2 to a plasma-resistant member is the Japanese Patent Application Laid-Open Publication No. 2012-206913 (the below-mentioned Patent Document 4). The Patent Document 4 discloses an invention wherein, since a sintered body of MgF2 simple has a defect of low mechanical strength, by mixing at least one kind of non-alkaline metallic dispersed particles having a lower mean linear thermal expansion coefficient than MgF2 such as Al2O3, AlN, SiC or MgO, the defect of low mechanical strength thereof can be compensated for. However, when a sintered body of such mixture is used as the above moderator to neutrons, the moderation performance thereof is greatly different from that of MgF2 simple because of the influence of the non-alkaline metal mixed into MgF2. Therefore, it is easily predicted that it is difficult to apply a sintered body of this kind of mixture to a use as a moderator. In addition, an example of an application of a sintered body mainly comprising CaF2 to a plasma-resistant member is the Japanese Patent Application Laid-Open Publication No. 2004-83362 (the below-mentioned Patent Document 5). The Patent Document 5 describes a method wherein using hydrofluoric acid, impurities other than Mg is removed from a low-purity raw material containing Mg, so as to precipitate high-purity CaF2, and a fluoride sintered body whose starting raw material is the high-purity CaF2 containing Mg of 50 ppm or more and 5% by weight or less is produced. A problem here is a state of Mg contained in the starting raw material, though the state is not described at all. And there is no description concerning the technique by which the degree of purity of the low-purity raw material is raised using hydrofluoric acid. Then, when presuming the process of raising the degree of purity of a low-purity raw material as a person skilled in the art, generally speaking, in the case of raising the degree of purity of a low-purity raw material using hydrofluoric acid, a method is often adopted, wherein impurities in the raw material are first dissolved into a hydrofluoric acid solution as many as possible, and if a component (Ca, here) desired to be a main raw material dissolved with impurities in this dissolution process, the component is precipitated and separated by use of the difference in solubility among the dissolved components. When further reviewing the invention, it is presumed that Mg exhibited different dissolution behavior from other impurities. In the specification, it is referred to as only “Mg”, and according to the descriptions of Examples in Table 1, as for high concentrations of impurity components other than Mg (such as Fe, Al, Na and Y), all of their concentrations were decreased by purity raising treatment, but only the concentration of Mg did not change, being 2000 ppm before the treatment and being also 2000 ppm after the treatment. Hence, there is a high possibility that Mg might be in a state which is hard to dissolve in hydrofluoric acid, that is, a metal state. If CaF2 containing metal Mg is a starting raw material, the sintering process thereof is very different from the case like the present invention wherein a mixture of CaF2 and MgF2 is a starting raw material, and the characteristics of the sintered bodies are also very different from each other. On the other hand, an invention relating to a neutron moderator was disclosed lately. That is the Japanese Patent No. 5112105 (Patent Document 6). The Patent Document 6 discloses ‘a moderator which moderates neutrons, comprising a first moderating layer obtained by melting a raw material containing calcium fluoride (CaF2), and a second moderating layer comprising metal aluminum (Al) or aluminum fluoride (AlF3), the first moderating layer and the second moderating layer being adjacent to each other’. In the Patent Document 6, the first moderating layer obtained by melting the raw material containing CaF2 is disclosed, but raw material conditions such as the purity, components, particle size and treatment method thereof, and melting conditions such as heating temperatures, holding times thereof and the type of heating furnace are not mentioned at all, very insincerely described as a patent specification. In the Patent Document 6, there is no description suggesting that something related to MgF2 should be used as a neutron moderator. In the preceding documents, as described above, there is no description suggesting the use of a sintered body of MgF2 as a moderator to neutrons, one kind of radiation. In such situation, the present inventors found that it was possible to use a MgF2 sintered body with a modification made thereto as a moderator to neutrons, one kind of radiation, and achieved the invention of the prior application. In the invention of the prior application, a high-purity MgF2 raw material is pulverized and two-stage compressing and molding step is conducted thereon. That is, after molding by a uniaxial press molding method, this press molded body is further molded by a cold isostatic pressing (CIP) method so as to form a CIP molded body. Then, by firing the same with three-level heating conditions using an atmosphere-adjustable normal pressure furnace, a sintered body having a compact structure is produced with suppressing foaming of MgF2 as much as possible. However, since MgF2 very easily generates foams, it is not easy to actually suppress its foaming. As a result, the range of relative densities (i.e. 100×[bulk density of a sintered body]/[true density](%)) of sintered bodies produced by this method was 92%-96%, and the mean value thereof was of the order of 94%-95%. A characteristic desired for a sintered body for a neutron moderator is ‘the mean value of relative densities of 95% or more at least, desirably 96% or more should be stably secured’. In order to achieve the characteristic, the present inventors worked toward further development and found that when using a MgF2—CaF2 binary system sintered body for radiation use, the relative density thereof could be easily improved, compared with a sintered body of MgF2 simple, and that by making the sintering conditions proper, sintered bodies having more desirable densities could be stably produced, leading to the completion of the present invention. Patent Document 1: Japanese Patent Application No. 2013-142704 (filed on Jul. 8, 2013) Patent Document 2: Japanese Patent Application Laid-Open Publication No. 2000-302553 Patent Document 3: Japanese Patent Application Laid-Open Publication No. 2000-86344 Patent Document 4: Japanese Patent Application Laid-Open Publication No. 2012-206913 Patent Document 5: Japanese Patent Application Laid-Open Publication No. 2004-83362 Patent Document 6: Japanese Patent No. 5112105 Non-Patent Document 1: H. Tanaka et al., Applied Radiation and Isotopes 69 (2011) 1642-1645 Non-Patent Document 2: H. Tanaka et al., Applied Radiation and Isotopes 69 (2011) 1646-1648 Non-Patent Document 3: Hiroaki Kumada, TetsuyaYamamoto, Dose Evaluation of Neutron Capture Therapy in JRR-4, Health Physics, 42(1), (2007) 23-37 The present invention was developed in order to solve the above problems, and it is an object of the present invention to provide a MgF2—CaF2 binary system sintered body for a radiation moderator, having excellent characteristics as a moderator used for moderating the energy of neutrons, a kind of radiation, in the good use of the neutrons for therapy, which makes it possible to enhance remedial values and downsize an apparatus for therapy, and which is inexpensive unlike a single crystal body, and a method for producing the same. It is another object of the present invention to provide a MgF2—CaF2 binary system sintered body for a radiation moderator having a very compact structure, without plane orientation dependency of the moderation performance originated in crystal orientation, unlike a single crystal body having such plane orientation dependency, and without ununiformity based on the structural defects such as subgrains, and a producing method by which such sintered bodies can be stably produced. The present inventors first gave basic consideration to the selection of substances (compounds) suitable for a moderator which performs shielding (i.e. moderation) to high-energy neutrons. That is, they first conducted relative research on the moderation performance to neutrons of substances of each kind. Here, whether or not neutrons moderated to intermediate energy in the wide energy range of neutrons can be moderated to an energy level with which a patient can be irradiated, was examined. As compounds to be examined, since the energy level of the injected neutrons was intermediate, compounds containing halogen elements, being compounds of relatively light elements were considered. Fluorides such as calcium fluoride (CaF2) and magnesium fluoride (MgF2), or chlorides such as calcium chloride (CaCl2) and magnesium chloride (MgCl2) were first conceived. Since chlorides easily generate molten salt (a liquid phase) at heating in producing the processed articles thereof, sintering reaction caused by use of the formation of a solid solution in which a solid phase and a liquid phase are mixed, is difficult to be caused therein. Even if a sintered body can be generated, it becomes chemically active and there is a high risk of lack of stability. Compared with chlorides, sintered bodies of fluorides are relatively chemically stable, and therefore, fluorides were selected in anticipation of superiority thereof over chlorides. As the basic characteristics required for a moderator other than moderation performance, a characteristic of keeping the shape of the product is exemplified. It is important to be excellent in mechanical strength with which damage in mechanical processing and in handling during manufacturing the product can be prevented. The mechanical strength of a sintered body is determined by micro strength of bonding parts between particles, the compactness of the sintered body, and moreover, the brittleness originated from a crystal structure (such as polycrystal or single crystal or amorphous) of the parent thereof. The compactness of the sintered body is determined by the defoaming state such as the sizes, shapes, distribution and number of bubbles, in other words, the shape such as the width and length of the bonding parts and a bound body (parent) of ex-particles. Basic technical ideas of the present invention are: (1) relaxing the sintering conditions by mixing two kinds of raw materials, that is, making it possible to conduct sintering at a lower temperature than the case of a raw material of one kind simple; (2) by regulating the particle size of the raw material provided to the sintering reaction, the particle growth by solid phase reaction is promoted, and a compact sintered body having strong cohesion between particles through the formation of a solid solution is formed; and (3) when a raw material of fluoride is heated at a high temperature, part of the raw material vaporizes (mainly part of the fluoride is thermally decomposed (sublimates) and generates fluorine gas), leading to the formation of bubbles (foaming). By sintering at a temperature low enough to avoid this foaming and making the process of heating (sintering heat pattern) proper, a compact sintered body is produced. By combining all the above technical ideas (1), (2) and (3), the present invention aims to stably produce fluoride sintered bodies of MgF2—CaF2 binary system for a radiation moderator excellent in moderation performance required as a member for a moderator to radiation, especially to neutrons and mechanical strength (shape keeping) characteristic as a fundamental characteristic other than the moderation performance. In the present invention, the MgF2—CaF2 binary system sintered body is mentioned mainly as a moderator to neutrons, but the sintered body has excellent performance as a member for shielding to not only neutrons but also other radioactive rays such as X-rays or gamma-rays. Concerning the technical idea (1), as shown in FIG. 4, each of the MgF2—CaF2 binary system sintered bodies within a range of good sintering conditions has a relative density, for example, the highest attained relative density tending to be higher by the order of 0.5%-1.5% than the sintered bodies of MgF2 simple. And the temperature limits of the secondary sintering temperature of the MgF2—CaF2 binary system sintered bodies, in which the bulk densities thereof become high, are wider, and it means that the stable sintering conditions can be easily satisfied. Concerning the technical idea (2), the effect of the particle size control of the raw material is described below. The chief object thereof is to make the particle size condition of the raw material proper so as to promote sintering by firing at a low temperature (hereinafter, referred to as ‘low-temperature sintering’). As the conditions of the raw material, each of high-purity raw material powders for a starting raw material (the mean particle diameter of the raw material powders was about 140 μm in median diameter) was pulverized by the below-mentioned pulverization method, and provided to the subsequent treatment step, and the characteristic evaluation of the completed sintered body was conducted. As a result, it was found that: (i) the particle size distribution range should be small, specifically, the maximum particle diameter should be 50 μm or less, desirably 30 μm or less; (ii) as a preferable state of the particle size distribution, the shape of the particle size distribution curve drawn with particle diameter (unit: μm) as the abscissa and particle diameter ratio (ratio of every particle diameter, unit: wt. %) as the ordinate, should be not ‘2-peak type’ or ‘3-peak type’, but ‘1-peak type’ or ‘sub-1-peak type’, when the shapes of the particle size distribution curves are divided into ‘1-peak type’, ‘sub-1-peak type’, ‘2-peak type’ and ‘3-peak type’ as the shapes of mountain ranges are expressed; (iii) the mean particle diameter should be 6 μm or less, desirably 4 μm or less in median diameter; and by simultaneously satisfying these conditions of the items (i), (ii) and (iii), the sintered body could be allowed to have a high density, leading to a noticeable improvement of moderation performance as a neutron moderator. The foaming phenomenon in the above technical idea (3) is described below. Using a differential thermal analyzer, alterations in weight and in endothermic and exothermic amount of the sample of the starting raw material were examined while heating. As a result, a minute quantity of weight decrease was found at approximately 800° C.-850° C., though there were slight differences depending on the mix proportion of the starting raw material. It appears that fluorine attached to a parent of a preliminary sintered body or fluorine resolving in the parent, for example, with a weak bonding property, dissociated and decomposed first of all. After further heating, a point of inflection of the weight decrease curve appeared at approximately 850° C.-900° C., and the weight decrease became noticeable. The results of this differential thermal analysis, and the examination results of the sintering conditions and the structure of the sintered body in a preliminary sintering test below-mentioned, specifically, the examination results such as 1. the situation of bubble generation in the sintered body, 2. the situation of tissue structure of the sintered portion, and 3. the bulk density of the sintered body were totally considered. It was anticipated that when heated at a temperature of the point of inflection of the weight decrease curve or higher, part of bonded fluorine element in MgF2 or CaF2 would start to decompose, and cause the generation of fluorine gas, leading to the formation of fine bubbles. Then, the temperature of this point of inflection, 850° C.-900° C. is referred to as the starting temperature of foaming (Tn). The temperature at which vaporization started was slightly different depending on the composition. In the case of a composition mainly comprising MgF2 (MgF2 of 70-99.8% by weight, and CaF2 of the rest), vaporization started at about 800° C. and it became quite brisk at about 850° C. (the point of inflection, that is, the starting temperature of foaming Tn was decided to be 850° C.). In the case of a composition mainly comprising CaF2 (MgF2 of 10-40% by weight, and CaF2 of the rest), vaporization started at about 850° C. and it became quite brisk at about 900° C. (similarly, Tn was decided to be 900° C.). In the case of MgF2 of 40-70% by weight and CaF2 of the rest, vaporization started at around the intermediate temperature between the above two cases, that is, in the temperature limits of about 825° C. or more, and it became quite brisk at about 875° C. (similarly, Tn was decided to be 875° C.). That is, in the case of the composition mainly comprising MgF2 (MgF2 of 70-99.8% by weight, and CaF2 of the rest), sublimation starts at about 800° C., and it becomes brisk and foaming starts at about 850° C. In the case of the composition mainly comprising CaF2 (MgF2 of 10-40% by weight, and CaF2 of the rest), sublimation starts at about 850° C. and it becomes brisk and foaming starts at about 900° C. In the case of MgF2 of 40-70% by weight, and CaF2 of the rest, the mix proportion intermediate therebetween, sublimation starts at about 825° C., and it becomes brisk and foaming starts at about 875° C. Thus, when a fluoride sublimates (a phenomenon in which a solid phase changes into a gas phase without passing through a liquid phase. In this case, a synonym for “vaporize”), fluorine gas is generated, resulting in generation of fine bubbles in the sintered body. The shapes of the generated bubbles are almost spheres. When observing the broken-cross section of the sintered body with an electron microscope (SEM), the cross sections of bubbles look like circles close to true circles. The sizes of the bubbles range from small ones of several μm to large ones of 20 μm-40 μm in diameter seen on the broken-cross section. The shapes of the small ones of several μm are approximately circles and the shapes of the large ones are rarely circles. Most of them are irregular such as long and narrow, or angular. Judging from these shapes, it is considered that the small ones are bubbles which have just been generated, and that the large ones are aggregates of some of the generated bubbles or residuals originated from voids between particles or the like which could not defoam in the sintering process. The reason why the value of relative density is shown by range corresponding to one value of bulk density is because in the case of a binary system sintered body of MgF2 and CaF2, the true densities of the both are different (that of MgF2 is 3.15 g/cm3, while that of CaF2 is 3.18 g/cm3), and therefore, depending on the mix proportion thereof, the true density of the mixture varies slightly. Here, the value of true density of the mixture is decided as shown below, so as to calculate the relative density thereof. It is decided that: (1) the true density is 3.15 g/cm3, in the case of a composition mainly comprising MgF2, that is, MgF2 of 70% by weight or more and 99.8% by weight or less (referred to as 70-99.8% by weight in the present application) and CaF2 of the rest; (2) the true density is 3.16 g/cm3, in the case of MgF2 of 40% by weight or more and less than 70% by weight (referred to as 40-70% by weight) and CaF2 of the rest; and (3) the true density is 3.17 g/cm3, in the case of MgF2 of 10% by weight or more and less than 40% by weight (referred to as 10-40% by weight) and CaF2 of the rest. In order to achieve the above object, a MgF2—CaF2 binary system sintered body for a radiation moderator according to a first aspect of the present invention is characterized by comprising MgF2 containing CaF2 from 0.2% by weight to 90% by weight inclusive, having a compact polycrystalline structure excellent in radiation moderation performance, especially neutron moderation performance with a bulk density of 2.96 g/cm3 or more. In the MgF2—CaF2 binary system sintered body for a radiation moderator according to the first aspect of the present invention, the difference between the parts of the organizational structure of the sintered body is small, and the generated quantity of melt is restricted and the crystal growth of a solid solution (a phase in which a solid phase and a liquid phase are mixed) is suppressed, leading to reducing the occurrence of brittle portions, resulting in enhanced strength of the sintered body. As a result, a sintered body having excellent moderation performance as a neutron moderator and enhanced mechanical strength, that is, all the characteristics required for a neutron moderator, can be obtained. The MgF2—CaF2 binary system sintered body for a radiation moderator according to a second aspect of the present invention is characterized by having a bending strength of 15 MPa or more and a Vickers hardness of 90 or more as regards mechanical strengths. The MgF2—CaF2 binary system sintered body for a radiation moderator according to the second aspect of the present invention has strong cohesion between particles, leading to high micro strength of the bonding part, noticeably improved mechanical strength. Accordingly, a sintered body having more excellent moderation performance as a neutron moderator and being extremely excellent in mechanical strength, can be provided. A method for producing a MgF2—CaF2 binary system sintered body for a radiation moderator according to a first aspect of the present invention is characterized by comprising the steps of: mixing a MgF2 powder with a CaF2 powder of 0.2-90% by weight and further adding 0.02-1% by weight of a sintering aid thereto to mix; molding the raw material powder compounded in the preceding step at a molding pressure of 5 MPa or more using a press molding device; molding the press molded article at a molding pressure of 5 MPa or more using a cold isostatic pressing (CIP) device; conducting preliminary sintering by heating the CIP molded article in a temperature range of 600° C.-700° C. in an air atmosphere; conducting sintering by heating the preliminary sintered body in a temperature range from (Tn-100)° C. to (Tn)° C. when the starting temperature of foaming of the preliminary sintered body is (Tn)° C., in an air atmosphere or in an inert gas atmosphere; and forming a sintered body having a compact structure by heating the same in a temperature range of 900° C.-1150° C. in the same atmosphere as the preceding step. According to the method for producing a MgF2—CaF2 binary system sintered body for a radiation moderator according to the first aspect of the present invention, the sintered body fired by this method has strong cohesion between particles, and high micro strength of the bonding part. The mechanical strength which was a problem to be solved is remarkably improved, and the sintered body can be used as a member for a neutron moderator without problems for actual use. The degree of compactness of the sintered body can be raised according to the selection of the mix proportion of MgF2—CaF2, heating atmosphere, heating temperature pattern and the like. The crystalline structure of the sintered body fired by this method is polycrystalline, resulting in remarkable improvement of the brittleness compared with a single crystal. And the highest attained relative density thereof within a range of good sintering conditions can be raised compared with a sintered body of MgF2 simple, and by making wider the temperature limits of the secondary sintering temperature thereof, in which the bulk density thereof becomes high, the stable sintering conditions can be easily realized. The method for producing a MgF2—CaF2 binary system sintered body for a radiation moderator according to a second aspect of the present invention is characterized by the shape of a particle size distribution curve of the compound which shows a sub-1-peak-type or 1-peak-type particle size distribution, wherein the maximum particle diameter is 50 μm or less and the median diameter of the particles is 6 μm or less in the method for producing a MgF2—CaF2 binary system sintered body for a radiation moderator according to the first aspect of the present invention. According to the method for producing a MgF2—CaF2 binary system sintered body for a radiation moderator according to the second aspect of the present invention, by more pulverizing the starting raw material and making the particle size range smaller so as to raise the packing density, the sintering reaction, especially low-temperature sintering which promotes cohesion of particles through solid phase reaction and bonding thereof can be promoted. Accordingly, the mechanical strength of the sintered body fired by this method can be remarkably enhanced. The method for producing a MgF2—CaF2 binary system sintered body for a radiation moderator according to a third aspect of the present invention is characterized by the inert gas atmosphere in the sintering step comprising one kind of gas or a mixture of plural kinds of gases, selected from among nitrogen, helium, argon and neon in the method for producing a MgF2—CaF2 binary system sintered body for a radiation moderator according to the first or second aspect of the present invention. By the method for producing a MgF2—CaF2 binary system sintered body for a radiation moderator according to the third aspect of the present invention, defoaming in the sintering process becomes easy to occur, and the relative density of the sintered body is easily raised. The preferred embodiments of the MgF2—CaF2 binary system sintered body for a radiation moderator having a compact polycrystalline structure excellent in radiation moderation performance, especially neutron moderation performance, and the method for producing the same according to the present invention are described below by reference to the Figures. In the method for producing a MgF2—CaF2 binary system sintered body according to the preferred embodiment, as shown in FIG. 3, a high-purity (purity of 99.9% by weight or more) MgF2 powder was mixed with a high-purity (purity of 99.9% by weight or more) CaF2 powder in the proportion of 0.2-90% by weight (included in a total of 100), and as a sintering aid, for example, a carboxymethyl cellulose (CMC) solution was added thereto in the proportion of 0.02-1% by weight (not included in 100) to 100 of the mixture and mixed. The mixture was used as a starting raw material (raw material mixing step). The starting raw material was molded at a molding pressure of 5 MPa or more using a uniaxial press device (uniaxial press molding step), and this press molded body was molded at a molding pressure of 5 MPa or more using a cold isostatic pressing (CIP) device (CIP molding step). Preliminary sintering was conducted by heating this CIP molded body in a temperature range of 600° C.-700° C. in an air atmosphere (preliminary sintering step). This preliminary sintered body was heated in a temperature range just below the starting temperature of foaming Tn, that is, in a temperature range from (Tn-100° C.) to Tn for a relatively long period of time (specifically, 3-12 hours) in an air atmosphere or in an inert gas atmosphere so as to allow sintering to make progress more uniformly (primary sintering step). The temperature range just below the starting temperature of foaming Tn was defined through the measurement using a differential thermal analyzer, and the temperature range varies in a range of about 750° C.-900° C. depending on the mix proportion of the raw materials of MgF2 and CaF2. As described above, it varies in a temperature range of 750° C.-850° C. in the case of a composition mainly comprising MgF2, in that of 800° C.-900° C. in the case of a composition mainly comprising CaF2, and in that of 775° C.-875° C. in the case of an intermediate composition of the both. Thereafter, in the same atmosphere, the same was heated in the vicinity of the temperature limits in which a solid solution starts to be formed (the temperature limits in the vicinity of 980° C., being a temperature at which a solid solution starts to be formed in the phase diagram of the MgF2—CaF2 binary system in FIG. 2), that is, in a temperature range of 900° C.-1150° C. for a relatively short period of time (0.5-8 hours), and then cooled so as to produce a MgF2—CaF2 binary system sintered body having a compact structure (secondary sintering step). The reason why the sintering step was divided into two steps, primary and secondary, is in order to suppress foaming as much as possible, and make the difference of the degree of sintering progress in every part (such as a periphery portion and a center portion) of the sintered body as small as possible. Particularly, in order to produce a large-size compact sintered body, the technique is important. The large size here is applied to press molded bodies in the below-described Examples having the size of about 220 mm×220 mm×H85 mm, while the small size is applied to the below-mentioned press molded bodies having the size of dia. 80 mm×H50 mm. In a test conducted in order to roughly grab proper heating conditions of the sintering step, the starting raw materials comprising MgF2—CaF2 binary system and MgF2 simple, respectively, were used, the sample size was the above large size, and both of the two stages of sintering were conducted in a nitrogen gas atmosphere. In the primary sintering, the temperature was held at 840° C. for 6 hours and in the subsequent secondary sintering, the heating time was set to be 2 hours with varied heating temperatures so as to measure the relative densities of the sintered bodies. As a result, as shown in FIG. 4, the relative densities of 95% or more could be secured in a wide range of heating conditions when two-stage sintering step was conducted, in either case of MgF2—CaF2 binary system and MgF2 simple, and particularly, in the case of MgF2—CaF2 binary system, those of 96%-97% could be obtained in the good condition range (heating at 950° C.-1050° C.). On the other hand, as shown in the below-described Comparative Examples 11 and 12, when only one-stage sintering step was conducted, the relative densities were 94% or less. The aim of mixing a CaF2 powder being a secondary raw material into a MgF2 powder being a main raw material is to cause the sintering reaction which allows the region of the formation of a solid solution on the phase diagram shown in FIG. 2 to become clearer, since MgF2 simple has a high melting point of 1252° C. and the temperature region of the formation of a solid solution is partially unclear, shown with dot lines. By mixing the right quantity of CaF2, being a fluoride of Ca which is presumed to have similar characteristics to Mg since Ca belongs to the same group as Mg on the periodic table of elements and its period is next to Mg, the melting point can be lowered and the temperature conditions of the formation of a solid solution can be clarified. By mixing CaF2, the melting point can be moved from the dot line region on the left end portion of the line indicating the temperature region of starting of the formation of a solid solution in FIG. 2 toward the solid line region of the intermediate mix proportions positioned on the right hand. As a result, it becomes easy to make the sintering temperature conditions proper. As a material to be mixed into MgF2 other than CaF2 being a fluoride of Ca, LiF being a fluoride of Li can be exemplified. As the sintering aid, two kinds, the CMC and the calcium stearate (SAC), were selected. With various adding proportions of each of them, the effects of addition thereof were confirmed. For comparison, a test with no sintering aid was also conducted. The main raw material MgF2 were mixed with the secondary raw material CaF2 in various mix proportions in a range of 0-97.5% by weight (included in a total of 100). After mixing using a ball mill for half a day, the two kinds of sintering aids were added in the proportion of 0-2% by weight (not included in the total), respectively. And using a pot mill, the same was mixed a whole day and night so as to obtain a starting raw material. The ball mill made of alumina having an inside diameter of 280 mm and a length of 400 mm was used, and balls of φ5: 1800 g, φ10: 1700 g, φ20: 3000 g and φ30: 2800 g, made of alumina were filled therein. The pot mill made of alumina having an inside diameter of 200 mm and a length of 250 mm was used. This compound of a prescribed quantity was filled into a wooden mold form, and using a uniaxial press device, compressed and molded at a uniaxial press pressure of 5 MPa or more. The inside size of the mold form used in the Examples was 220 mm×220 mm×H150 mm, and the inside size of the mold form used in a small-size test was 80 mm in diameter and 100 mm in height. This press molded body was put into a thick vinyl bag, which was then deaired and sealed, and it was put through a cold isostatic pressing (CIP) device. The press molded body was put into a molding part having a two-split structure (inside diameter 350 mm×H120 mm), which was sealed. The space between the vinyl bag with the press molded body inside and the molding part was filled with clean water, and then, isostatic pressing was conducted at a hydraulic pressure of 5 MPa or more so as to form a CIP molded body. The preliminary sintering step was conducted on the CIP molded bodies in an air atmosphere with various kinds of conditions in a heating temperature range of 500° C. to 750° C. and in a heating time range of 3 to 18 hours. After observing the appearance of the preliminary sintered bodies, the preliminary sintered bodies were sintered with the conditions which were regarded as good sintering conditions in the preceding preliminary test. The sintering step was conducted with the conditions wherein, in a nitrogen gas atmosphere, the temperature was raised from room temperature to 600° C. at a fixed rate for 6 hours, and held there for 8 hours, and then, it was raised to 1000° C. at a fixed rate for 2 hours and held there for 1 hour. And thereafter, it was lowered to 100° C. for 20 hours. By observing the appearance of the taken-out sintered bodies, the state of compactness of the inside thereof and the like, proper raw material mix proportions, raw material processing conditions, preliminary sintering conditions and the like were investigated. As a result, in cases where the mix proportion of the secondary raw material CaF2 to the main raw material MgF2 was less than 0.2% by weight, the sintering performance did not become much better due to mixing of CaF2. The difference in compactness between the inside portion and the periphery portion of the sintered body was likely to be large as is the case with MgF2 simple. Therefore, in order to improve the sintering performance by mixing thereof, it was judged that CaF2 of 0.2% by weight or more was required. On the other hand, in the case of 90.1% by weight or more, a larger number of large bubbles were left in the inside portion of the sintered body, compared with the periphery portion thereof, resulting in insufficient compactness. Judging from these situations, the mix proportions of CaF2 to MgF2, in which the difference in compactness between the inside portion and the periphery portion of the sintered body was small, that is, the sintering performance was in a good state, were 0.2-90% by weight. It was confirmed that the more desirable mix proportions thereof in which the difference in compactness between the inside portion and the periphery portion of the sintered body was smaller, resulting in an excellent degree of uniformity, were 1.5-80% by weight. Hence, the proper range of mix proportions of CaF2 was judged to be 0.2-90% by weight, more desirably 1.5-80% by weight. There was no big difference between the effects of the two kinds of sintering aids, but when the mix proportion of the sintering aid was less than 0.02% by weight, the shape keeping performance of the molded body was poor. And when the mix proportion thereof exceeded 1.1% by weight, coloring which appeared to be a residual of the sintering aid was noticed on the preliminary sintered body or the sintered body in some cases. Hence, the proper range of mix proportions of the sintering aid was judged to be 0.02-1% by weight. In a uniaxial press test using the above wooden mold form for a small-size test, when the molding pressure of the uniaxial press device was less than 5 MPa, the press molded body easily lost its shape in handling. As the molding pressure was gradually increased from 5 MPa, the bulk density of the press molded body gradually increased, and it was recognized that the bulk densities of the preliminary sintered body and the sintered body also tended to increase though slightly. The test was conducted with the molding pressure gradually increased to 100 MPa. However, even if the molding pressure was raised to 20 MPa or more, no improvement of performance of the preliminary sintered body or the sintered body was recognized. Hence, the proper value of the molding pressure of the uniaxial press device was decided to be 5 MPa or more, desirably 20 MPa. When the molding pressure of the CIP device was less than 5 MPa, the CIP molded body easily lost its shape in handling. As the molding pressure was gradually increased from 5 MPa, the bulk density of the CIP molded body gradually increased, and it was recognized that the bulk densities of the preliminary sintered body and the sintered body also tended to increase though slightly. The test was conducted with the CIP molding pressure gradually increased to 60 MPa. However, even if the molding pressure was raised to 20 MPa or more, no great improvement of performance of the preliminary sintered body or the sintered body was recognized. Hence, the proper value of the molding pressure of the CIP device was decided to be 5 MPa or more, desirably 20 MPa. The research of preliminary sintering conditions of the CIP molded body in an air atmosphere was conducted under the below-described conditions. By mixing MgF2 with CaF2 of 3% by weight, and adding CMC of 0.1% by weight as a sintering aid thereto, a starting raw material was prepared. Using the wooden mold form for a small-size test, by setting the molding pressure of a uniaxial press device to be 20 MPa and setting the molding pressure of a CIP device to be 20 MPa, CIP molded bodies were formed. Using the CIP molded bodies formed under such conditions, the preliminary sintering conditions were researched. At heating temperatures of less than 600° C., shrinkage was small compared with the size of the molded body, while at heating temperatures of 710° C. or more, the shrinkage rate was too high and therefore, shrinkage was difficult to control. Hence, the proper range of the preliminary sintering temperatures was decided to be 600° C.-700° C. Concerning the heating time, at 600° C., it was judged that 8-9 hours were optimal, and that 4-10 hours were proper, judging from the evaluation of the shrinkage rate. At 700° C., it was judged that 6-8 hours were optimal, and that 4-10 hours were proper. From these results, the heating conditions in the preliminary sintering step were decided to be at 600° C.-700° C. for 4-10 hours in an air atmosphere. What is likely to give most influence on the performance of a sintered body in producing the MgF2—CaF2 binary system sintered body for a radiation moderator is the sintering step. From the above researches and tests, the proper conditions until just before the sintering step were clarified. The sintering step and the sintering mechanism which appear to be desirable to a MgF2—CaF2 binary system sintered body for a radiation moderator are put in order. The terms “primary flocculation process” and “secondary flocculation process” which express the degrees of progress of the sintering step, are described below. The “primary flocculation process” refers to an event in the first half of the stage of sintering, and in the initial stage thereof, the intervals between particles gradually become narrower and the voids among particles also become smaller. With further progress of sintering, the particle-to-particle contact portions become thick and the voids among them become further smaller. Here, the majority of the voids are open pores connecting to the surrounding atmosphere. Up to this stage is called “primary flocculation process”. On the other hand, after the end of the primary flocculation process, with further progress of sintering, the open pores gradually decrease and turn into closed pores. Roughly, the stage of turning into closed pores and the subsequent stage of defoaming and compacting are generically called “secondary flocculation process”. In the producing method according to the preferred embodiment, due to raw material mixing, particle size control, mixing, two-stage molding (uniaxial press molding and CIP molding), preliminary sintering and the like, it was noticed that the voids among particles of the preliminary sintered body were small, and that the voids almost uniformly scattered without gathering (the first half stage of the primary flocculation process). In the heating process of the next sintering step, the heating temperature is gradually raised. Around the temperature limits (500° C.-550° C.) slightly lower than the preliminary sintering temperatures (600° C.-700° C.), particles start to gather, and thereafter, solid phase reaction starts in the temperature limits far lower than 980° C. at which a solid solution starts to be formed. With that, flocculation of particles makes progress, leading to shorter particle-to-particle distances and smaller voids. It is generally said that the solid phase reaction starts in the temperature limits lower by the order of 10% or further lower than the temperature at which a solid solution starts to be formed. From the observation results in the preliminary test by the present inventors, it was considered that the solid phase reaction started in far lower temperature limits than the above generally said temperature limits, in the order of 500° C.-550° C. It can be said on the ground that at 600° C., the lowest limit of the preliminary sintering temperature, sintering by the solid phase reaction has already made progress considerably so that the preliminary sintered body considerably shrinks compared with the CIP molded body. It is considered that the solid phase reaction makes progress at a low reaction rate in the temperature limits and that it makes progress at a quite high reaction rate in the temperature limits in the vicinity of 750° C., or more up to 980° C. Here, in the case of heating at relatively low temperatures (600° C.-700° C.) like assumed preliminary sintering for a short period of time, most of the voids remain in a state of open pore (which is the first half stage of the primary flocculation process). What attention should be paid to here is behavior of fine bubbles (foaming bubbles) generated through vaporization of part of the raw material in the temperature limits of about 850° C.-900° C. or more, as mentioned above. In the case of heating at about 1000° C. or more, the heating time should be as short as possible, since this formation of foaming bubbles comes to be noticeable. In the producing method according to the preferred embodiment, the sintering step is divided into two. In the primary sintering step, by heating in the relatively low temperature limits in which no foaming bubbles are formed for a long period of time, sintering of the whole body is allowed to make progress almost uniformly. The micro structure of the sintered body comprises mainly open pores, but part of them is turned into closed pores (after finishing the second half stage of the primary flocculation process, partially in the secondary flocculation process). In the secondary sintering step, heating is conducted in the relatively high temperature limits in the vicinity of 980° C. at which a solid solution starts to be formed for a minimum required period of time. As the micro structure of the sintered body, the formation of foaming bubbles is suppressed as much as possible, while the sintering reaction is allowed to make progress so as to turn almost all the open pores into closed pores, that is, the secondary flocculation process is finished, resulting in obtaining a high-density sintered body. Micro behavior of raw material particles is described here. It is presumed that particles of CaF2 are present around particles of the main raw material MgF2 and promote interface reaction with the particles of MgF2. Around a heating temperature exceeding 980° C. at which a solid solution starts to be formed, melting starts in the vicinity of a particle interface where the particles of CaF2 are present, and a solid solution of a MgF2—CaF2 binary system compound starts to be formed. It is presumed that this solid solution fills the voids among particles and that in some part, finer voids are also filled therewith through capillary phenomenon. On the other hand, even if the heating temperature is lower than 980° C., by heating and holding at about 750° C. or more for a relatively long period of time as described above, the solid phase reaction easily makes progress, the voids gradually decrease with the elapse of time so as to be closed pores. Parallel with that, a gas component within the closed pores diffuses within the bulk (parent) of the sintered body, leading to the progress of defoaming so as to make the sintered body compact with few bubbles (which is the secondary flocculation process). Also here, in heating at temperatures not lower than the starting temperature of foaming Tn (the starting temperature of foaming differs depending on the mix proportion of the raw materials MgF2 and CaF2), that is, temperatures exceeding 850° C.-900° C., attention should be paid to the formation of fine bubbles (foaming bubbles) generated through vaporization of the raw material. That is because it is presumed that the foaming bubbles contain fluorine gas, and it is considered that this gas is a relatively heavy element and difficult to diffuse in the bulk of the sintered body. As measures for that, to avoid heating in the temperature limits of vaporization as much as possible, and if necessary, to heat at a temperature as low as possible or to heat for a short period of time are considered. The difference in appearance between such foaming bubbles and bubbles left after pores became closed but could not be defoamed in the sintering step (hereinafter, referred to as residual bubbles) is described below. The sizes of the foaming bubbles generated by general heating for a relatively short period of time are approximately several μm in diameter, and the shapes thereof are almost perfect spheres. On the other hand, the shapes of the residual bubbles are not perfect spheres but irregular, and the sizes thereof are all mixed up, large, medium and small. Therefore, it is possible to distinguish the both according to the difference in shape. Here, in the case of high-temperature heating at temperatures far exceeding 1160° C., or heating at temperatures exceeding 1160° C. for a long period of time, a foaming bubble and a foaming bubble, or a residual bubble and a foaming bubble gather and grow to a large irregular bubble in some cases, resulting in difficulty in judging its origin. With the progress of the secondary flocculation process, the voids among particles become smaller, and all or most of the voids are surrounded by particles or a bridge portion of the sintered body so as to be closed pores (bubbles). Or depending on the conditions, gases are released through the voids (open pores), or gases within the bubbles permeate into the bulk (parent) such as the particles or the bridge portion of the sintered body to degas, resulting in no bubbles (referred to as a defoaming phenomenon). Whether the voids among particles are left as closed pores, that is, bubbles, or by degassing, no bubbles are formed, is a significant element for deciding the degree of achievement of compactness of the sintered body, leading to deciding the characteristics of the sintered body. Particularly in the case of sintering in a light element gas atmosphere such as He or Ne among inert gases, it is considered that the lighter element more easily diffuses within the pores or the bulk of the sintered body, leading to promoting the capillary phenomenon and defoaming phenomenon, so that bubbles are difficult to remain, leading to easy compacting. Thus, in order to make the whole compact, it is important to advance the primary flocculation process (in detail, it is presumed that the primary flocculation process is divided into the first half stage and the second half stage) and the secondary flocculation process almost simultaneously and almost uniformly on the whole in each process. In the invention according to the preferred embodiment, the preliminary sintering step chiefly equivalent to the first half stage of the primary flocculation process, the primary sintering step chiefly equivalent to the second half stage of the primary flocculation process, and the secondary sintering step chiefly equivalent to the secondary flocculation process are separately conducted, so as to make the two flocculation processes easy to make progress almost uniformly throughout the sintered body. However, Even if the sintering step is divided into two steps of preliminary sintering and sintering like this, a noticeable difference in degree of compactness is caused without proper heating conditions. For example, in the case of heating at high temperatures exceeding the proper limits in the preliminary sintering step, in the case of rapidly heating at the temperature raising stage of the sintering step, or in cases where the holding temperature in the sintering step is a high temperature exceeding the proper limits, a remarkable difference in degree of compactness is caused between the periphery portion and the inside portion of the sintered body. By improper heating, degassing becomes difficult in the process of compacting of the inside portion of the sintered body, and the compactness of the inside portion thereof is likely to be insufficient. It means that it is important to make the heating temperature pattern in the sintering step proper according to the size. Particularly, when producing a large-size sintered body, it is necessary to strictly control the heating conditions since a difference in degree of compactness between the periphery portion and the inside portion of such sintered body is easily caused. In order to clarify the relationship between the sample size and the sintering state, the present inventors conducted a small-size test using samples molded in a mold form of a uniaxial press device the inside size of 80 mm in diameter and 100 mm in height, and a large-size test using samples molded in a mold form thereof the inside size of 220 mm×220 mm×H150 mm. As a result, in the small-size test, there were cases where a high-density sintered body having a relative density exceeding 95% was obtained depending on the heating conditions even if one sintering step was conducted. On the other hand, in the large-size test, with one sintering step, any of the sintered bodies had a low density of less than 94% under the same sintering conditions as the small-size test. What is important here is that the whole of the preliminary sintered body has already advanced almost uniformly to the first half stage of the primary flocculation. Only preliminary sintered bodies in a state in which the whole body has already advanced to the first half stage of the primary flocculation were provided to these tests of the sintering step. The description of the sintering step test A mixture of a main raw material MgF2 with CaF2 of 3% by weight, and a raw material of MgF2 simple as a comparative material were used as starting materials. CMC of 0.1% by weight was added thereto as a sintering aid. And using the above mold form for a large-size test, the compounds were molded at a molding pressure of 20 MPa of a uniaxial press device and at a molding pressure of 20 MPa of a CIP device. The CIP molded bodies were preliminary sintered at 650° C. for 6 hours in an air atmosphere so as to obtain preliminary sintered bodies. In a nitrogen atmosphere, as primary sintering step, the preliminary sintered bodies were heated to 840° C. and the temperature was held there for 6 hours and then raised to a secondary sintering temperature for 2 hours. The secondary sintering temperature was varied from 700° C. to 1250° C., at an interval of every 50° C., and the temperatures each were held for 2 hours. Thereafter, the heating was stopped and the temperature was lowered by self-cooling (so-called furnace cooling) for about 20 hours, and when reaching 100° C. or lower at which time it was previously set to take out the sintered body, it was taken out. As a result of the sintering test with such two-stage sintering step, as shown in FIG. 3, in the case of a range from 900° C. to 1150° C., most of the bulk densities of the sintered bodies exceeded 2.96 g/cm3, which were high. The true density of the binary system compound was 3.15 g/cm3 and the relative density thereof was 94.0%, while the true density of the raw material of MgF2 simple was also 3.15 g/cm3 and the relative density thereof was also 94.0%. In either case of sintering temperatures of less than 900° C., and those of 1160° C. or more, the relative densities were lower than 94.0% (the bulk density of 2.96 g/cm3). The sintered bodies of the MgF2—CaF2 binary system raw material tended to have a higher relative density by the order of 0.5%-1.5% than those of MgF2 simple in a range of good sintering conditions. When observing the sections of those sintered bodies, in the case of sintered bodies sintered at temperatures lower than 900° C., not many but some open pores were noticed in some of them, wherein the bridge width of the sintered portion was narrow, so that it could be regarded as absolutely insufficient progress of sintering. In the case of sintered bodies sintered at temperatures of 1160° C. or more, especially 1200° C. or more, those had a porous pumiceous structure as if bubbles were innumerably formed inside. Fine bubbles which were almost perfect spheres of several to dozen μm in diameter were observed all over the sintered body and innumerable irregular bubbles (foaming bubbles and aggregates thereof) of 10 μm or more in diameter were found all over the sections. From another examination using a differential thermal analyzer by the present inventors, it was found out that when heating the compound of MgF2—CaF2 binary system, the weight clearly started to decrease at a temperature of about 800° C.-850° C. (the temperature becomes gradually higher within the temperature range as the mix proportion of CaF2 to MgF2 increases), and that the weight started to drastically decrease at about 850° C.-900° C. This means that a sublimation phenomenon in which MgF2 or CaF2 starts to dissolve/vaporize to generate fluorine starts due to heating at about 800° C.-850° C. or more. A foaming phenomenon through this fluorine sublimation is noticeably caused by heating at about 850° C.-900° C. or more, and fine bubbles are formed all over the sintered body. The behavior of the foaming bubbles such as defoaming or remaining as bubbles is decided according to the degree of progress of the sintering step, in which portion of the sintered body they were formed and the like. In the primary flocculation process, for example, since the whole sintered body contains mainly open pores, the majority of foaming bubbles can be defoamed through the open pores, leading to few bubbles left. In the secondary flocculation process, since the sintered body contains mainly closed pores, a large number of foaming bubbles cannot be defoamed, leading to remaining as bubbles. To swiftly complete the sintering in the secondary flocculation process leads to suppressing foaming and reducing residual bubbles. Hence, it is preferable that the transition from the primary flocculation process to the secondary flocculation process should be advanced in the whole sintered body with as small a difference of the degree of progress as possible among the portions thereof. However, it is not easy to undergo the transition from the primary flocculation process to the secondary flocculation process in the whole sintered body without a difference of the degree of progress among the portions thereof. Then, the present inventors considered the below-described method. Heating at a relatively low temperature in the temperature limits just below the starting temperature of foaming Tn (850° C.-900° C.), specifically in the temperature limits between (Tn-100° C.) and Tn for a relatively long period of time was conducted, so that the primary flocculation process and the first half of the secondary flocculation process were completed. And then, by heating at a temperature in the vicinity of the temperature (980° C.) at which a solid solution starts to be formed for a relatively short period of time, the second half of the secondary flocculation process was completed. By such sintering, the degree of progress of sintering could be made uniform in the whole sintered body, and the formation of bubbles could be suppressed as much as possible. How the proper sintering conditions were decided is described below. In the same manner as the above sintering condition change test, a main raw material MgF2 was mixed with CaF2 of 3% by weight. CMC of 0.1% by weight was added thereto as a sintering aid. The same was molded using a mold form for a large-size test at a molding pressure of 20 MPa of a uniaxial press device and a molding pressure of 20 MPa of a CIP device. Preliminary sintering was conducted on this CIP molded body at 650° C. which was held for 6 hours in an air atmosphere. As the conditions of the sintering step, the atmosphere was set to be a nitrogen gas atmosphere. Preliminary tests concerning each of heating and cooling conditions in the heating pattern were conducted in three cases of the required time of 4, 6 and 8 hours. As a result, in the case of 4 hours, small cracks occurred in the sintered body, while in the other cases, the results were good. Therefore, the required time was set to be 6 hours, shorter one selected from 6 and 8 hours. The atmosphere was set to be a nitrogen gas atmosphere, and the heating temperature was varied in a range of 700° C. to 1250° C. In eleven cases of the holding time of 2, 3, 4, 5, 6, 8, 10, 12, 14, 16 and 18 hours, the tests were conducted. As a result, in the case of less than 750° C., the compactness was insufficient, regardless of the holding time. In the case of heating at 750° C., the compactness was insufficient with a holding time of 4 hours or less. On the other hand, in the case of heating at 1160° C. or more, a large number of bubbles were generated due to too fast sintering speed, regardless of the holding time. In the case of a holding time of 18 hours, in some cases, foaming occurred in part of the periphery of the sintered body, leading to getting out of shape in appearance. Reviewing the results, in the case of heating at 750° C., the sintering state was good with a holding time of 14 and 16 hours. In the case of heating at 800° C., the sintering state was good with a holding time of 10 and 12 hours, while slightly insufficient with 6 and 8 hours, and beyond decision of quality with 14 hours or more. In the case of 830° C., the sintering state was good with a holding time of 10 and 12 hours. In the case of 850° C., the sintering state was good with a holding time of 8, 10 and 12 hours, while slightly insufficient with 5 hours, and beyond decision of quality with 14 hours or more. In the case of 900° C., the sintering state was good with a holding time of 5 to 12 hours, while slightly insufficient with 4 hours, and beyond decision of quality with 14 hours or more. In the case of 1000° C., the sintering state was good with a holding time of 5 to 12 hours, while slightly insufficient with 4 hours, and beyond decision of quality with 14 hours or more. In the case of 1050° C., the sintering state was good with a holding time of 5 to 10 hours, while slightly insufficient with 4 hours, and much foaming was seen with 12 hours or more. In the case of 1100° C., the sintering state was good with a holding time of 4 to 8 hours, while slightly insufficient with 3 hours or less, and much foaming was seen with 10 hours or more. In the case of 1150° C., the sintering state was good with a holding time of 2 and 3 hours, while much foaming was seen with 4 hours or more. In the case of 1160° C. or more, much foaming was seen with any holding time, and the results were beyond decision of quality or poor because of too much melting. Here, when the heating temperature was a comparatively low temperature of 750° C. to 850° C., the sintering state was good with a holding time of 6 to 12 hours, while that was slightly insufficient with a holding time of 3 to 5 hours. Since the method according to the preferred embodiment has the subsequent secondary sintering step, with the evaluation in this step (equivalent to the primary sintering step), the holding time of 3-12 hours was regarded as a good heating condition. In order to examine the relationship between the heating temperature and the bulk density of the sintered body, using the same preliminary sintered bodies as the above, the heating temperature was varied within a range of 600° C. to 1300° C. (with a holding time of 6 hours in any case). As a result, in the case of a heating temperature of 850° C., the bulk density was approximately 2.96 g/cm3 (the relative density of 94.0%). The sintered body having a bulk density of that value or more was judged to have sufficient compactness without troubles such as losing its shape in the treatment of the second step. On the other hand, in the case of heating temperatures of 1160° C. or more, in some cases, foaming occurred in part of the periphery of the sintered body, resulting in a trouble such as getting out of shape in appearance. From the above examination results of the sintering conditions and the relationship between the heating temperature and the bulk density, it was judged that, if the sintering step was one heating step, the heating temperature of 850° C. to 1150° C. and the holding time of 3 to 12 hours were proper. What was clarified here is, when relatively long time heating, such as at 900° C. for 14 hours or more, at 1000° C. for 14 hours or more, at 1100° C. for 10 hours or more, or at 1150° C. for 8 hours or more, was conducted, a large number of foaming bubbles were generated and part of those gathered and grew to large bubbles. Such sintered body involved defects which would cause cracks to occur from a large bubble portion or cause splitting in processing of the next mechanical processing step. From these situations, as a fundamental plan of the sintering step, it was decided that foaming should be suppressed as much as possible, and the sintering reaction should be allowed to sufficiently make progress, leading to producing a sintered body having a good processability in the subsequent mechanical processing step. At the first stage of the sintering step (the primary sintering step), it was aimed to suppress foaming to a minimum, to allow the sintering to make slow progress, and to minimize a difference of the degree of progress between the inside portion and the periphery portion of the sintered body. Therefore, the heating temperature was decided to be within the above range of 700° C. to 1150° C. Since the starting temperature of foaming Tn is 850° C. in the case of a raw material mainly comprising MgF2, it was judged that the heating temperature should be 850° C. or less, not exceeding the temperature. On the other hand, since the sintering state was insufficient in the case of heating at temperatures lower than the Tn by 100° C. or more, it was judged that the heating temperature at the first stage of the sintering step should be between (Tn-100° C.) and Tn, between 750° C. and 850° C. in the case of a raw material mainly comprising MgF2. The proper heating conditions in the primary sintering step were the heating temperature between (Tn-100° C.) and Tn, and the holding time of 3-12 hours. The same tendency was found even in cases where the mix proportion of CaF2 to MgF2 varied between 0.5-90% by weight. Heating at the stage of enhancing the sintering reaction of the sintered body, that is, heating in the secondary sintering step, was decided to be conducted properly in the temperature limits in the vicinity of 980° C. at which a solid solution starts to be formed, that is, 900° C. to 1150° C. It was aimed to make the holding time as short as possible in order to enhance the sintering reaction and suppress foaming. The proper holding time was decided to be 0.5 to 8 hours, since the enhancement of the sintering reaction was poor in the case of less than 0.5 hour, and too many bubbles were formed in the case of 9 hours or more. The examination of the proper conditions of the heating temperature and the holding time in the secondary sintering process when the atmospheric gas was changed from nitrogen gas to helium gas is described below. A mixture of a main raw material MgF2 with CaF2 of 3% by weight was used as a starting material, to which CMC of 0.1% by weight was added as a sintering aid. Using a mold form of press molding for a large-size test, the material was molded at a molding pressure of 20 MPa of a uniaxial press device and at a molding pressure of 20 MPa of a CIP device. This CIP molded body was preliminary sintered at 650° C. for 6 hours in an air atmosphere so as to obtain a preliminary sintered body. Using helium gas as the atmospheric gas in the primary and secondary sintering processes, the preliminary sintered body was heated to 840° C. which was held for 6 hours as primary sintering. Then, it was raised to each of secondary sintering temperatures varying in a range of 700° C. to 1250° C., at an interval of every 50° C. for 2 hours, and the target temperature was held for 2 hours. And then, the heating was stopped and the temperature was lowered by self-cooling (so-called furnace cooling) for about 20 hours, and when reached a predetermined taking-out temperature of 100° C. or lower, the sintered body was taken out. As a result of the sintering test with the above two-stage sintering step, as shown in FIG. 5, in the case of a temperature range of 900° C. to 1150° C., most of the sintered bodies had a high bulk density exceeding 2.96 g/cm3. The true density of the binary system compound was 3.15 g/cm3 and the relative density thereof was 94.0%, while the true density of the raw material of MgF2 simple was also 3.15 g/cm3 and the relative density thereof was also 94.0%. In either case of sintering temperatures lower than 900° C., and those of 1160° C. or more, the relative density was lower than 94.0% (the bulk density of 2.96 g/cm3). The sintered bodies sintered in a helium gas atmosphere tended to have a higher relative density within a range of good sintering conditions by the order of 0.5%-1% than in a nitrogen gas atmosphere. It is considered that the reason why the bulk density becomes high in a helium gas atmosphere is because the diffusion velocity of helium gas within the bulk (parent) of the sintered body is higher than that of nitrogen gas. It is presumed that, since helium gas more easily diffuses within the bulk than nitrogen gas, when voids become closed pores with the progress of sintering in the sintering process, part of the closed pores disappear without becoming bubbles, or the sizes of the closed pores become smaller. However, helium gas showed better effects within a range of the above proper sintering conditions, while the effects were not all-around, being not noticeably seen in the region other than the proper sintering conditions. As the reasons of such result, it was considered that under the sintering conditions outside the proper range, for example, there was a limit in improving too slow sintering speed due to an insufficient heating condition, or in the case of an excessive heating condition, the ununiformity of the sintering speed of every part of the sintered body could not be improved by enhancing the diffusivity of helium gas in the bulk. In the case of helium gas, when the heating temperature in the sintering step was less than 900° C., regardless of the holding time, or in the case of a holding time of 4 hours or less, the compactness was insufficient. When the heating temperature was 1160° C. or more, the sintering speed was too high, regardless of the holding time, as is the case with nitrogen gas, resulting in occurrence of a large number of bubbles, and in the case of a holding time of 16 hours or more, because of foaming, the appearance got out of shape in some cases. Accordingly, in the case of a starting raw material made by mixing mainly MgF2 with CaF2, it was judged that the proper range of sintering temperatures was 900° C.-1150° C., regardless of the kind of inert atmospheric gas in the sintering step. Furthermore, in the case of sintering temperatures of 930° C.-1100° C., even when the sintered body was provided to the mechanical processing, structural defects such as cracks were difficult to occur, leading to good mechanical processability. As a result, it was judged that the sintering temperature was more preferably in a temperature range of 930° C.-1100° C. Therefore, as proper heating conditions of the sintering step in a helium gas atmosphere, as is the case with the above nitrogen gas atmosphere, the proper condition of the primary sintering step was in a range of 750° C. or more and less than the starting temperature of foaming, while that of the secondary sintering step was in a temperature range of 900° C.-1150° C. The inert gas is not limited to nitrogen and helium. In the case of argon or neon, the same effects can be obtained. Moreover, since neon is expected to have high solubility or high diffusivity in the parent of the sintered body, like helium, the defoaming phenomenon can be more promoted and effects equal to those of helium can be expected. When the heating conditions of the sintering step were within the proper range, the state of the completed sintered body was wholly compact in any case, and no clearly defective portion such as a locally-found large void or a crack seen in a general ceramic sintered body could be found in this sintered body. As the particle size control of a MgF2 powder and a CaF2 powder each, using a container of a pot mill made of alumina the size of an inside diameter of 200 mm and a length of 250 mm as a ball mill, balls made of alumina, φ20 mm: 3000 g and φ30 mm: 2800 g, were filled therein. And about 3000 g of each of the raw material powders was filled therein and rotated for a prescribed period of time. The rotation was stopped every two or three days so as to take powder samples and measure the same. The particle size distribution were measured using ‘a laser diffraction particle size analyzer (type number: SALD-2000)’ made by Shimadzu Corporation according to JIS R 1629 ‘Determination of particle size distributions for fine ceramic raw powders by laser diffraction method’. The sample preparation at that time was conducted according to JIS R1622 ‘General rules for the sample preparation of particle size analysis of fine ceramic raw powder’. As the light source of the SALD-2000, a semiconductor laser of a wavelength of 680 nm is used. The sensitivity to particles having a diameter larger than this wavelength (about 1 μm or more) was good and the measurement accuracy was high. On the other hand, as for the sensitivity to fine particles of the order of submicron, it was considered that the measurement accuracy was low compared with the particles having a large diameter though some way to improve the measurement accuracy was devised. Therefore, it is considered that the actual number of fine particles of the order of submicron may be larger than the analysis result. In other words, ‘there is a high possibility that the ratio of fine particles may be larger than the analysis result in the actual particle size distribution, and that the mean particle diameter may be smaller than the shown value thereof.’ However, in the present application, the values of the particle sizes measured according to the above measurement method are shown as they are. FIG. 1 shows a particle size distribution in the case of the above pulverization of a MgF2 raw material. It was found that the median diameter was about 10 μm after three-day pulverization, about 8 μm after five-day pulverization, about 6 μm after one-week pulverization, about 5 μm after two-week pulverization, and about 3 μm after four-week pulverization. Even if a MgF2 raw material and a CaF2 raw material were mixed, the particle size distribution similar to that of the raw material of MgF2 simple could be obtained. Concerning the original raw material powder and the above particles whose mean particle diameter became about 4 μm after three-week pulverization, the shapes in appearance of the particles of them, respectively, were observed using an electron microscope. In the original raw material powder, some irregular-shaped particles, mainly angular particles were seen, while most of the particles after three-week pulverization were rounded. It was found that most of the angular portions of the particles of the original raw material powder were worn by pulverization so as to be approximated to sphere shapes. The shape of the particle size distribution curve of the powder after this particle size control can be expressed by being likened to the shape of a mountain range. When the shape of the curve looks like as if “two peaks” or “three peaks” run in a line, it is called ‘2-peak type’ or ‘3-peak type’. The curve of three-day pulverization and that of five-day pulverization obviously showed a high ratio of coarse particle portions, respectively, which were regarded as ‘2-peak type’ or ‘3-peak type’. On the other hand, in the case of one-week pulverization and two-week pulverization, respectively, the ratio of the coarse particle portions substantially decreased, the coarse particle portion of 30 μm or more remained several % by weight, but that of 50 μm or more almost disappeared, and the shape of the particle size distribution curve was reaching almost the 1-peak type having a small gently inclined portion around the particle diameter of 10 μm-15 μm (this state is called ‘sub-1-peak type’). And in the case of four-week pulverization, the coarse particle portion of 30 μm or more almost disappeared and the shape of the particle size distribution curve could be approximately similar to a normal distribution (this state is called ‘1-peak type’). Thus, the particle shapes were rounded and approached sphere shapes by pulverization of the raw material powder, and the ratio of coarse particles decreased, resulting in a great change of the shape of the particle size distribution curve from ‘2-peak type’ or ‘3-peak type’ to ‘sub-1-peak type’, and further to ‘1-peak type’. This change exerted a noticeably good influence on sintering reaction in the sintering process. Examples according to the present invention are described below by reference to the Figures, but the present invention is not limited to the below-described Examples. Here, a characteristic evaluation test conducted on sintered bodies is described. Samples for evaluation were prepared by prototyping large-size sintered bodies (rough size of the sintered body: about 205 mm×about 205 mm×H about 70 mm) and conducting mechanical processing such as cut-out in the shape of a required sample thereon. In order to evaluate the neutron moderation performance, as shown in the above Non-Patent Documents 1 and 2, a beam emitted from an accelerator was allowed to collide with a plate made of Be being a target, and by nuclear reaction, high-energy neutrons (fast neutrons) were mainly generated. Then, using Pb and Fe each having a large inelastic scattering cross section as a moderator in the first half of moderation, the fast neutrons were moderated to some extent (approximately, up to 1 MeV) while suppressing the attenuation of the number of neutrons. The moderated neutrons were irradiated to a moderator to be evaluated (a moderator in the second half of moderation), and by examining the neutrons after moderation, the moderator was evaluated. The examination of the neutrons was conducted according to the method described in the above ‘Non-Patent Document 3’. The moderators to be evaluated were made of raw materials MgF2 and CaF2 in some varied mix proportions. Through the mixing step of each kind of raw materials, molding step and sintering step, a high-density MgF2—CaF2 binary system sintered body having a relative density in a fixed range (95.0±0.5%) was produced. The total thickness of a moderator in the second half was set to be 600 mm in any case. What was evaluated here is the dose of epithermal neutrons having intermediate-level energy which is effective for therapy, and how many fast neutrons and gamma-rays having high-level energy which has a high possibility of adversely influencing a patient (side effects), remained in the neutrons moderated by the moderator. The results are shown in FIG. 6 (Table 1). The dose of epithermal neutrons effective for therapy slightly varied, as the quantity of CaF2 mixed into MgF2 was increased, but the digit of the neutron flux (dose) of epithermal neutrons was the ninth power in any case, so that regardless of the mix proportion, the dose thereof sufficient for therapy was secured. On the other hand, the mix rate of fast neutrons having a high possibility of adversely influencing a patient (the ratio of fast neutron dose in the total neutron dose after passing through a moderator) was the lowest in the case of mixing CaF2 of several to 10% by weight. It gradually increased as the mix proportion thereof far exceeded these mix proportions and increased to 20% by weight, and to 40% by weight. It was the highest when CaF2 was 100% by weight. The mix rate of gamma-rays having the next highest possibility of adversely influencing a patient after fast neutrons (the ratio of gamma-ray dose in the total neutron dose after passing through a moderator) was a low digit of E−14 (the minus 14th power), regardless of the mix proportion of CaF2 to MgF2. The influence of gamma-rays was small, regardless of the mix proportion of CaF2. From these results, it was proved that when the main raw material MgF2 was mixed with CaF2 of 2-10% by weight, it had the most excellent performance as a moderator. Even if the mix proportion was other than such mix proportions, for example, 0.2% by weight or more and less than 2% by weight, or 10.1% by weight or more and 90% by weight or less, the neutrons were on the level usable for therapy. The evaluation results are limited to the cases where the relative density of the sintered body is roughly within a fixed range (95.0±0.5%). The higher relative density the sintered body has, the lower the residual dose of fast neutrons is. Conversely, the lower relative density the sintered body has, the higher the residual dose of fast neutrons is. Accordingly, the importance of improvement of the density of the sintered body is the same. Concerning the moderation performance of a moderator to neutrons, it was sufficient only that a MgF2—CaF2 binary system sintered body having a compact structure should have a bulk density of 2.96 g/cm3 or more. The moderator to neutrons is required to have mechanical strength other than moderation performance. It was proved by the below-described examination of mechanical strength that the sintered body for a radiation moderator according to the present invention had sufficient mechanical strength, with which it could be used without problems in processing and molding such as cutting-off, grinding, polishing, cleaning and drying as a moderating member in a moderation system device for BNCT, and further in handling such as the installation thereof in the moderation system device. Even if it was irradiated with neutrons, it was capable of resisting their irradiation impacts, being extremely excellent. As mechanical strengths, bending strength and Vickers hardness were examined. The samples for bending strength, having a size of 4 mm×46 mm×t3 mm with the upper and lower surfaces optically polished were prepared according to JIS C2141, and tested according to the three-point bending test JIS R1601. To obtain the Vickers hardness, according to JIS Z2251-1992, using ‘Micro Hardness Tester’ made by Shimadzu Corporation, an indenter having a load of 100 g was pressed for 5 seconds of loading time so as to measure the diagonal length of the impression, which was converted into hardness in the following manner.Vickers hardness=0.18909×P/(d)2 Here, P: load (N) and d: diagonal length of impression (mm) A high-purity MgF2 powder being a main raw material (mean particle diameter of 4 μm and purity of 99.9% by weight or more) was mixed with a CaF2 powder (mean particle diameter of 4 μm and purity of 99.9% by weight or more) of 1.5% by weight, and mixed using a ball mill for 12 hours. Thereafter, a carboxymethyl cellulose (CMC) solution was added thereto as a sintering aid in the proportion of 0.1% by weight to 100 of the mixture, which was mixed in a pot mill for 12 hours so as to obtain a starting raw material. This starting raw material was filled into a mold form (inside size of 220 mm×220 mm×H150 mm) of a uniaxial press device and compressed at a uniaxial press pressure of 20 MPa to be molded. This press molded body (size of about 220 mm×220 mm×t85 mm), which was put into a thick vinyl bag and sealed after deairing, was put into a molding part (inside size: dia. 350 mm×H120 mm) of a cold isostatic pressing (CIP) device. Clean water was filled into the space between the vinyl bag with the press molded body therein and the CIP molding part, and by isostatic pressing at a molding pressure of 20 MPa at room temperature, CIP molding was conducted. The preliminary sintering step was conducted on this CIP molded body at 650° C. for 6 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 800° C. at a fixed rate for 6 hours in a nitrogen gas atmosphere, and the temperature was held there for 8 hours. It was then raised to 1050° C. at a fixed rate for 2 hours and held there for 1.5 hours. Heating was then stopped, and the temperature was lowered by self-cooling (so-called furnace cooling) for about 20 hours to 100° C. or less at which time it was previously set to take out the sintered body, after which it was taken out. The bulk density of the sintered body was calculated at 3.02 g/cm3 (the true density of this compound is 3.15 g/cm3 and the relative density thereof is 95.9%, and hereinafter, referred to as “true density of 3.15 g/cm3 and relative density of 95.9%”) from the bulk volume of the appearance thereof and the weight thereof. The sintering state thereof was good. Since the appearance of the sintered body was a square form, the ‘bulk density’ here was obtained by a method wherein the bulk volume was calculated from the measured two sides of the square and thickness, and the weight separately measured was divided by the bulk volume. This also applied to the following. Using a sample taken from this sintered body, evaluation tests of neutron moderation performance and characteristics of every kind were conducted. The results are shown in FIG. 7 (Table 2). This also applied to the following Examples and Comparative Examples. Here, concerning a sintered body of MgF2 simple and a sintered body of CaF2 simple, being comparative materials, the neutron moderation performance and mechanical strengths were measured like the Examples and Comparative Examples. The sintered body in Example 1 showed excellent neutron moderation performance, and the mechanical strengths thereof were also good enough not to cause problems in handling in the next step. A MgF2 powder being a main raw material (mean particle diameter of 6 μm and purity of 99.9% by weight) was mixed with a CaF2 powder (mean particle diameter of 6 μm and purity of 99.9% by weight) of 0.2% by weight, and mixed using a ball mill for 12 hours. Thereafter, with the same molding conditions as in the above Example 1, a CIP molded body was produced, and the preliminary sintering step was conducted on this CIP molded body at 640° C. for 6 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 800° C. at a fixed rate for 6 hours in a helium gas atmosphere, and the temperature was held there for 5 hours. It was then raised to 920° C. at a fixed rate for 4 hours and held there for 1 hour. Then, the temperature was lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. The bulk density of the sintered body was 2.97 g/cm3 (true density of 3.15 g/cm3 and relative density of 94.3%). It was rather light, but the sintering state thereof was not unusual in appearance. Any of the evaluation results of the neutron moderation performance and mechanical strengths were good as shown in Table 2. The same MgF2 powder was mixed with the same CaF2 powder of 2% by weight as those in the above Example 1 and mixed using a ball mill for 12 hours. Thereafter, calcium stearate (SAC) of 1.0% by weight was added thereto as a sintering aid. The compound was mixed in a pot mill for 12 hours so as to obtain a starting raw material. Using a uniaxial press device, the press molding was conducted at a press pressure of 30 MPa, and then, using a cold isostatic pressing (CIP) device, the CIP molding was conducted at a CIP pressure of 30 MPa. The preliminary sintering step was conducted on this CIP molded body at 700° C. for 6 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 840° C. at a fixed rate for 6 hours in an air atmosphere, and the temperature was held there for 8 hours. It was then raised to 1150° C. at a fixed rate for 2 hours and held there for 0.75 hour. The temperature was then lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. The bulk density of the sintered body was 3.06 g/cm3 (true density of 3.15 g/cm3 and relative density of 97.1%), and the sintering state thereof was good. Any of the evaluation results of the neutron moderation performance and mechanical strengths were good as shown in Table 2. The same MgF2 powder was mixed with the same CaF2 powder of 3% by weight as those in the above Example 1 and mixed using a ball mill for 12 hours. Thereafter, a CMC solution of 0.03% by weight was added thereto as a sintering aid. The compound was mixed in a pot mill for 12 hours so as to obtain a starting raw material. Using a uniaxial press device, the press molding was conducted at a press pressure of 30 MPa, and then, using a cold isostatic pressing (CIP) device, the CIP molding was conducted at a CIP pressure of 30 MPa. The preliminary sintering step was conducted on this CIP molded body at 660° C. for 8 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 830° C. at a fixed rate for 6 hours in a nitrogen gas atmosphere, and the temperature was held there for 6 hours. It was then raised to 1080° C. at a fixed rate for 2 hours and held there for 2 hours. The temperature was then lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. The bulk density of the sintered body was 3.07 g/cm3 (true density of 3.15 g/cm3 and relative density of 97.5%), and the sintering state thereof was good. Any of the evaluation results of the neutron moderation performance and mechanical strengths were good as shown in Table 2. The same MgF2 powder was mixed with the same CaF2 powder of 7.5% by weight as those in the above Example 1 and mixed using a ball mill for 12 hours. Thereafter, calcium stearate (SAC) of 0.07% by weight was added thereto as a sintering aid. The compound was mixed in a pot mill for 12 hours so as to obtain a starting raw material. Using a uniaxial press device, the press molding was conducted at a press pressure of 40 MPa, and then, using a cold isostatic pressing (CIP) device, the CIP molding was conducted at a CIP pressure of 40 MPa. The preliminary sintering step was conducted on this CIP molded body at 690° C. for 8 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 830° C. at a fixed rate for 6 hours in a nitrogen gas atmosphere, and the temperature was held there for 9 hours. It was then raised to 1080° C. at a fixed rate for 2 hours and held there for 2 hours. The temperature was then lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. The bulk density of the sintered body was 3.06 g/cm3 (true density of 3.15 g/cm3 and relative density of 97.1%), and the sintering state thereof was good. Any of the evaluation results of the neutron moderation performance and mechanical strengths were good as shown in Table 2. A MgF2 powder being a main raw material (mean particle diameter of 5 μm and purity of 99.9% by weight) was mixed with a CaF2 powder (mean particle diameter of 5 μm and purity of 99.9% by weight) of 18% by weight, and mixed using a ball mill for 12 hours. Thereafter, a CMC solution of 0.3% by weight was added thereto as a sintering aid. The compound was mixed in a pot mill for 12 hours so as to obtain a starting raw material. Using a uniaxial press device, the press molding was conducted at a press pressure of 6 MPa, and then, using a cold isostatic pressing (CIP) device, the CIP molding was conducted at a CIP pressure of 15 MPa. The preliminary sintering step was conducted on this CIP molded body at 630° C. for 8 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 820° C. at a fixed rate for 6 hours in a nitrogen gas atmosphere, and the temperature was held there for 6 hours. It was then raised to 930° C. at a fixed rate for 2 hours and held there for 4 hours. The temperature was then lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. The bulk density of the sintered body was 2.98 g/cm3 (true density of 3.15 g/cm3 and relative density of 94.6%), and the sintering state thereof was good. Any of the evaluation results of the neutron moderation performance and mechanical strengths were good as shown in Table 2. The same MgF2 powder was mixed with the same CaF2 powder of 2.5% by weight as those in the above Example 6 and mixed using a ball mill for 12 hours. Thereafter, a CMC solution of 0.1% by weight was added thereto as a sintering aid. The compound was mixed in a pot mill for 12 hours so as to obtain a starting raw material. Using a uniaxial press device, the press molding was conducted at a press pressure of 30 MPa, and then, using a cold isostatic pressing (CIP) device, the CIP molding was conducted at a CIP pressure of 30 MPa. The preliminary sintering step was conducted on this CIP molded body at 650° C. for 6 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 840° C. at a fixed rate for 6 hours in a nitrogen gas atmosphere, and the temperature was held there for 5 hours. It was then raised to 1150° C. at a fixed rate for 2 hours and held there for 0.5 hours. The temperature was then lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. The bulk density of the sintered body was 3.01 g/cm3 (true density of 3.15 g/cm3 and relative density of 95.6%), and the sintering state thereof was good. Any of the evaluation results of the neutron moderation performance and mechanical strengths were good as shown in Table 2. The same MgF2 powder was mixed with the same CaF2 powder of 50% by weight as those in the above Example 1 and mixed using a ball mill for 12 hours. Thereafter, a CMC solution of 1% by weight was added thereto as a sintering aid. The compound was mixed in a pot mill for 12 hours so as to obtain a starting raw material. Using a uniaxial press device, the press molding was conducted at a press pressure of 7 MPa, and then, using a cold isostatic pressing (CIP) device, the CIP molding was conducted at a CIP pressure of 12 MPa. The preliminary sintering step was conducted on this CIP molded body at 600° C. for 5 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 860° C. at a fixed rate for 6 hours in a helium gas atmosphere, and the temperature was held there for 8 hours. It was then raised to 1080° C. at a fixed rate for 2 hours and held there for 2 hours. The temperature was then lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. The bulk density of the sintered body was 3.02 g/cm3 (true density of 3.16 g/cm3 and relative density of 95.6%). It was rather light, but the sintering state thereof was not unusual in appearance. Any of the evaluation results of the neutron moderation performance and mechanical strengths were good as shown in Table 2. The same MgF2 powder was mixed with the same CaF2 powder of 50% by weight as those in the above Example 2 and mixed using a ball mill for 12 hours. Thereafter, a CMC solution of 1% by weight was added thereto as a sintering aid. The compound was mixed in a pot mill for 12 hours so as to obtain a starting raw material. Using a uniaxial press device, the press molding was conducted at a press pressure of 30 MPa, and then, using a cold isostatic pressing (CIP) device, the CIP molding was conducted at a CIP pressure of 30 MPa. The preliminary sintering step was conducted on this CIP molded body at 610° C. for 7 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 860° C. at a fixed rate for 6 hours in a nitrogen gas atmosphere, and the temperature was held there for 8 hours. It was then raised to 970° C. at a fixed rate for 2 hours and held there for 4 hours. The temperature was then lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. The bulk density of the sintered body was 3.00 g/cm3 (true density of 3.16 g/cm3 and relative density of 94.9%). It was somewhat light, but the sintering state thereof was not unusual in appearance. Any of the evaluation results of the neutron moderation performance and mechanical strengths were good as shown in Table 2. The same MgF2 powder was mixed with the same CaF2 powder of 75% by weight as those in the above Example 1 and mixed using a ball mill for 12 hours. Thereafter, SAC of 0.07% by weight was added thereto as a sintering aid. The compound was mixed in a pot mill for 12 hours so as to obtain a starting raw material. Using a uniaxial press device, the press molding was conducted at a press pressure of 8 MPa, and then, using a cold isostatic pressing (CIP) device, the CIP molding was conducted at a CIP pressure of 10 MPa. The preliminary sintering step was conducted on this CIP molded body at 650° C. for 5 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 880° C. at a fixed rate for 6 hours in a nitrogen gas atmosphere, and the temperature was held there for 8 hours. It was then raised to 1060° C. at a fixed rate for 2 hours and held there for 3 hours. The temperature was then lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. The bulk density of the sintered body was 3.02 g/cm3 (true density of 3.17 g/cm3 and relative density of 95.3%). It was somewhat light, but the sintering state thereof was not unusual in appearance. Any of the evaluation results of the neutron moderation performance and mechanical strengths were good as shown in Table 2. The same MgF2 powder was mixed with the same CaF2 powder of 88% by weight as those in the above Example 6 and mixed using a ball mill for 12 hours. Thereafter, a CMC solution of 1% by weight was added thereto as a sintering aid. The compound was mixed in a pot mill for 12 hours so as to obtain a starting raw material. Using a uniaxial press device, the press molding was conducted at a press pressure of 30 MPa, and then, using a cold isostatic pressing (CIP) device, the CIP molding was conducted at a CIP pressure of 30 MPa. The preliminary sintering step was conducted on this CIP molded body at 650° C. for 5 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 880° C. at a fixed rate for 6 hours in a nitrogen gas atmosphere, and the temperature was held there for 8 hours. It was then raised to 950° C. at a fixed rate for 2 hours and held there for 4 hours. The temperature was then lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. The bulk density of the sintered body was 3.01 g/cm3 (true density of 3.17 g/cm3 and relative density of 95.0%). It was somewhat light, but the sintering state thereof was not unusual in appearance. Any of the evaluation results of the neutron moderation performance and mechanical strengths were good as shown in Table 2. The same MgF2 powder was mixed with the same CaF2 powder of 88% by weight as those in the above Example 1 and mixed using a ball mill for 12 hours. Thereafter, a CMC solution of 1% by weight was added thereto as a sintering aid. The compound was mixed in a pot mill for 12 hours so as to obtain a starting raw material. Using a uniaxial press device, the press molding was conducted at a press pressure of 8 MPa, and then, using a cold isostatic pressing (CIP) device, the CIP molding was conducted at a CIP pressure of 10 MPa. The preliminary sintering step was conducted on this CIP molded body at 650° C. for 5 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 880° C. at a fixed rate for 6 hours in a helium gas atmosphere, and the temperature was held there for 8 hours. It was then raised to 1120° C. at a fixed rate for 2 hours and held there for 2 hours. The temperature was then lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. The bulk density of the sintered body was 3.04 g/cm3 (true density of 3.17 g/cm3 and relative density of 95.9%), and the sintering state thereof was not unusual in appearance. Any of the evaluation results of the neutron moderation performance and mechanical strengths were good as shown in Table 2. A MgF2 powder being a main raw material (mean particle diameter of 8 μm and purity of 99.9% by weight) was mixed with a CaF2 powder (mean particle diameter of 8 μm and purity of 99.9% by weight) of 1.5% by weight, and mixed using a ball mill for 12 hours. Thereafter, SAC of 0.07% by weight was added thereto as a sintering aid. The compound was mixed in a pot mill for 12 hours so as to obtain a starting raw material. In a similar manner to the above Example 1, using a uniaxial press device, the press molding was conducted at a press pressure of 20 MPa, and then, using a cold isostatic pressing (CIP) device, the CIP molding was conducted at a CIP pressure of 20 MPa. The preliminary sintering step was conducted on this CIP molded body at 550° C. for 8 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 670° C. at a fixed rate for 6 hours in a nitrogen gas atmosphere, and the temperature was held there for 6 hours. It was then raised to 1200° C. at a fixed rate for 2 hours and held there for 2 hours. The temperature was then lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. The bulk density of the sintered body was 2.93 g/cm3 (true density of 3.15 g/cm3 and relative density of 93.0%), which was light. When observing the inside of the sintered body, there were a myriad of large bubbles of 0.1 mm or more in diameter. It was considered that these large bubbles were aggregates of fine foaming bubbles, or those of foaming bubbles and residual bubbles since the high mix proportion of 12% by weight of the MgF2 powder having a low melting point and heating at a high temperature of 1200° C. in the last sintering step allowed foaming to easily occur. Some insufficient levels of neutron moderation performance and mechanical strengths were noticed. A MgF2 powder being a main raw material (mean particle diameter of 10 μm and purity of 99.9% by weight) was mixed with a CaF2 powder (mean particle diameter of 10 μm and purity of 99.9% by weight) of 0.2% by weight, and a starting raw material was prepared in a similar manner to the Comparative Example 1. Using a uniaxial press device, the press molding was conducted at a press pressure of 4 MPa, and then, the CIP molding was conducted on this press molded body at a molding pressure of 4 MPa so as to obtain a CIP molded body in a similar manner to the Example 1. The preliminary sintering step was conducted on this CIP molded body at 600° C. for 6 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 830° C. at a fixed rate for 6 hours in a nitrogen gas atmosphere, and the temperature was held there for 5 hours. It was then raised to 950° C. at a fixed rate for 2 hours and held there for 4 hours. The temperature was then lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. The bulk density of the sintered body was 2.90 g/cm3 (true density of 3.15 g/cm3 and relative density of 92.1%), which was relatively light. This sintered body was soaked in pure water colored with a small quantity of ink solution for about 1 hour, and after raising it therefrom, the broken-cross section thereof was observed. The periphery portion thereof was wholly colored with this ink solution. It was considered that due to insufficient sintering, a large number of open pores were left. Some insufficient levels of neutron moderation performance and mechanical strengths were noticed. A MgF2 powder being a main raw material (mean particle diameter of 12 μm and purity of 99.9% by weight) was mixed with a CaF2 powder (mean particle diameter of 12 μm and purity of 99.9% by weight) of 5% by weight, and mixed using a ball mill for 12 hours. Thereafter, a CMC solution of 1.0% by weight was added thereto as a sintering aid. The compound was mixed in a pot mill for 12 hours so as to obtain a starting raw material. In a similar manner to the Example 1, using a uniaxial press device, the press molding was conducted at a press pressure of 20 MPa, and then, the CIP molding was conducted on this press molded body at a molding pressure of 20 MPa so as to obtain a CIP molded body. The preliminary sintering step was conducted on this CIP molded body at 700° C. for 10 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 900° C. at a fixed rate for 6 hours in a nitrogen gas atmosphere, and the temperature was held there for 10 hours. It was then raised to 1200° C. at a fixed rate for 2 hours and held there for 4 hours. The temperature was then lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. In part of the periphery portion of the sintered body, peeling was noticed. It was considered that this peeling was caused since foaming bubbles and residual bubbles gathered in the periphery portion and part of the periphery portion was cracked by the internal pressure of the bubbles. Here, since some part of the sintered body lost its shape, the bulk density thereof could not be measured. The same MgF2 powder was mixed with the same CaF2 powder of 5% by weight as those in the above Comparative Example 1, and a starting raw material was prepared in a similar manner thereto. Using a uniaxial press device, the press molding was conducted at a press pressure of 3 MPa, and then, this press molded body was CIP molded at a molding pressure of 3 MPa so as to obtain a CIP molded body in a similar manner to the above Example 1. The preliminary sintering step was conducted on this CIP molded body at 600° C. for 6 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 900° C. at a fixed rate for 6 hours in a nitrogen gas atmosphere, and the temperature was held there for 5 hours. It was then raised to 1200° C. at a fixed rate for 2 hours and held there for 2 hours. The temperature was then lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. Since there was a broken part in the periphery edge portion of the sintered body, the obtained bulk density thereof was an approximate value of about 2.92 g/cm3 (true density of 3.15 g/cm3 and relative density of 92.7%). Some insufficient levels of neutron moderation performance and mechanical strengths were noticed. The same MgF2 powder was mixed with the same CaF2 powder of 25% by weight as those in the above Comparative Example 1, and a starting raw material was prepared in a similar manner thereto. Using a uniaxial press device, the press molding was conducted at a press pressure of 30 MPa, and then, this press molded body was CIP molded at a molding pressure of 30 MPa so as to obtain a CIP molded body in a similar manner to the Example 1. The preliminary sintering step was conducted on this CIP molded body at 550° C. for 8 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 870° C. at a fixed rate for 6 hours in a nitrogen gas atmosphere, and the temperature was held there for 6 hours. It was then raised to 1160° C. at a fixed rate for 2 hours and held there for 3 hours. The temperature was then lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. The bulk density of the sintered body was 2.93 g/cm3 (true density of 3.15 g/cm3 and relative density of 93.0%). Some insufficient levels of neutron moderation performance and mechanical strengths were noticed. The same MgF2 powder was mixed with the same CaF2 powder of 25% by weight as those in the above Comparative Example 1, and a starting raw material was prepared in a similar manner thereto. Using a uniaxial press device, the press molding was conducted at a press pressure of 4 MPa, and then, this press molded body was CIP molded at a molding pressure of 4 MPa so as to obtain a CIP molded body in a similar manner to the above Example 1. The preliminary sintering step was conducted on this CIP molded body at 600° C. for 6 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 830° C. at a fixed rate for 6 hours in a nitrogen gas atmosphere, and the temperature was held there for 5 hours. It was then raised to 950° C. at a fixed rate for 2 hours and held there for 4 hours. The temperature was then lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. The bulk density of the sintered body was 2.91 g/cm3 (true density of 3.15 g/cm3 and relative density of 92.4%). Some insufficient levels of neutron moderation performance and mechanical strengths were noticed. The same MgF2 powder was mixed with the same CaF2 powder of 50% by weight as those in the above Comparative Example 1, and a starting raw material was prepared in a similar manner thereto. Using a uniaxial press device, the press molding was conducted at a press pressure of 20 MPa, and then, this press molded body was CIP molded at a molding pressure of 20 MPa so as to obtain a CIP molded body in a similar manner to the Example 1. The preliminary sintering step was conducted on this CIP molded body at 550° C. for 8 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 880° C. at a fixed rate for 6 hours in a nitrogen gas atmosphere, and the temperature was held there for 5 hours. It was then raised to 1200° C. at a fixed rate for 2 hours and held there for 2 hours. The temperature was then lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. The bulk density of the sintered body was 2.91 g/cm3 (true density of 3.16 g/cm3 and relative density of 92.1%). Some insufficient levels of neutron moderation performance and mechanical strengths were noticed. The same MgF2 powder was mixed with the same CaF2 powder of 50% by weight as those in the above Comparative Example 1, and a starting raw material was prepared in a similar manner thereto. Using a uniaxial press device, the press molding was conducted at a press pressure of 4 MPa, and then, this press molded body was CIP molded at a molding pressure of 4 MPa so as to obtain a CIP molded body in a similar manner to the Example 1. The preliminary sintering step was conducted on this CIP molded body at 600° C. for 6 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 850° C. at a fixed rate for 6 hours in a nitrogen gas atmosphere, and the temperature was held there for 5 hours. It was then raised to 960° C. at a fixed rate for 2 hours and held there for 4 hours. The temperature was then lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. The bulk density of the sintered body was 2.92 g/cm3 (true density of 3.16 g/cm3 and relative density of 92.4%). Some insufficient levels of neutron moderation performance and mechanical strengths were noticed. The same MgF2 powder was mixed with the same CaF2 powder of 88% by weight as those in the above Comparative Example 1, and a starting raw material was prepared in a similar manner thereto. Using a uniaxial press device, the press molding was conducted at a press pressure of 20 MPa, and then, this press molded body was CIP molded at a molding pressure of 20 MPa so as to obtain a CIP molded body in a similar manner to the Example 1. The preliminary sintering step was conducted on this CIP molded body at 530° C. for 8 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 900° C. at a fixed rate for 6 hours in a nitrogen gas atmosphere, and the temperature was held there for 6 hours. It was then raised to 1160° C. at a fixed rate for 2 hours and held there for 4 hours. The temperature was then lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. The bulk density of the sintered body was 2.90 g/cm3 (true density of 3.17 g/cm3 and relative density of 91.5%). Some insufficient levels of neutron moderation performance and mechanical strengths were recognized. The same MgF2 powder was mixed with the same CaF2 powder of 88% by weight as those in the above Comparative Example 1, and a starting raw material was prepared in a similar manner thereto. Using a uniaxial press device, the press molding was conducted at a press pressure of 4 MPa, and then, this press molded body was CIP molded at a molding pressure of 4 MPa so as to obtain a CIP molded body in a similar manner to the Example 1. The preliminary sintering step was conducted on this CIP molded body at 600° C. for 6 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 860° C. at a fixed rate for 6 hours in a nitrogen gas atmosphere, and the temperature was held there for 6 hours. It was then raised to 970° C. at a fixed rate for 2 hours and held there for 5 hours. The temperature was then lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. The bulk density of the sintered body was 2.93 g/cm3 (true density of 3.17 g/cm3 and relative density of 92.4%). Some insufficient levels of neutron moderation performance and mechanical strengths were noticed. The same MgF2 powder was mixed with the same CaF2 powder of 3% by weight as those in the above Comparative Example 1, and a starting raw material was prepared in a similar manner thereto. Using a uniaxial press device, the press molding was conducted at a press pressure of 30 MPa, and then, this press molded body was CIP molded at a molding pressure of 30 MPa so as to obtain a CIP molded body. The preliminary sintering step was conducted on this CIP molded body at 660° C. for 8 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 1060° C. at a fixed rate for 6 hours in a nitrogen gas atmosphere, and the temperature was held there for 2 hours. The temperature was then lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. The sintering step was conducted only in one stage (only the secondary sintering of the primary and secondary sintering was conducted). The bulk density of the sintered body was 2.93 g/cm3 (true density of 3.15 g/cm3 and relative density of 93.0%). Some insufficient levels of neutron moderation performance and mechanical strengths were noticed. The same MgF2 powder was mixed with the same CaF2 powder of 25% by weight as those in the above Comparative Example 1, and a starting raw material was prepared in a similar manner thereto. Using a uniaxial press device, the press molding was conducted at a press pressure of 30 MPa, and then, this press molded body was CIP molded at a molding pressure of 30 MPa so as to obtain a CIP molded body. The preliminary sintering step was conducted on this CIP molded body at 650° C. for 6 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 1150° C. at a fixed rate for 6 hours in a nitrogen gas atmosphere, and the temperature was held there for 1.5 hours. The temperature was then lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. The sintering step was conducted only in one stage (only the secondary sintering of the primary and secondary sintering was conducted). The bulk density of the sintered body was 2.90 g/cm3 (true density of 3.15 g/cm3 and relative density of 92.1%). Some insufficient levels of neutron moderation performance and mechanical strengths were noticed. [Comparative Material 1] Using the same powder of MgF2 simple as that in the above Example 6, a starting raw material was prepared in a similar manner to the Example 1. In a similar manner to the Example 1, using a uniaxial press device, the press molding was conducted at a press pressure of 20 MPa, and then, this press molded body was CIP molded at a molding pressure of 20 MPa so as to obtain a CIP molded body. The preliminary sintering step was conducted on this CIP molded body at 600° C. for 6 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 840° C. at a fixed rate for 6 hours in a nitrogen gas atmosphere, and the temperature was held there for 6 hours. It was then raised to 1100° C. at a fixed rate for 2 hours and held there for 2 hours. The temperature was then lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. The bulk density of the sintered body was 2.97 g/cm3 (true density of 3.15 g/cm3 and relative density of 94.3%). The neutron moderation performance of the sintered body was good enough to compare favorably with the Examples as shown in Table 2. On the other hand, the mechanical strengths thereof were within a range of good levels as shown in Table 2, but equivalent to the lower levels of strengths in the Examples. For information, this Comparative Material 1 is equivalent to a sintered body according to the prior application. [Comparative Material 2] Using the same CaF2 powder being a secondary raw material as that in the Example 6, a starting raw material was prepared in a similar manner to the Example 6. In a similar manner to the Example 1, using a uniaxial press device, the press molding was conducted at a press pressure of 20 MPa, and then, this press molded body was CIP molded at a molding pressure of 20 MPa so as to obtain a CIP molded body. The preliminary sintering step was conducted on this CIP molded body at 600° C. for 6 hours in an air atmosphere so as to obtain a preliminary sintered body. This preliminary sintered body was heated from room temperature to 880° C. at a fixed rate for 6 hours in a nitrogen gas atmosphere, and the temperature was held there for 6 hours. It was then raised to 1130° C. at a fixed rate for 2 hours and held there for 2 hours. The temperature was then lowered by furnace cooling to a predetermined taking-out temperature of 100° C., and the sintered body was taken out. The bulk density of the sintered body was 3.00 g/cm3 (true density of 3.18 g/cm3 and relative density of 94.3%). The mechanical strengths of the sintered body were good, while some insufficient levels of neutron moderation performance were noticed. |
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abstract | Methods are disclosed for determining a reticle pattern to be defined on a reticle used for charged-particle-beam microlithography performed using a high beam-acceleration voltage. The pattern is determined so as to define pattern elements, destined for transfer-exposure to respective edges of chips, on the reticle in a manner serving to reduce proximity effects in such elements when imprinted on the substrate, whether or not the elements are in peripherally situated chips (located at or near a wafer perimeter) or in chips located centrally on the substrate. On the reticle the profile of such an element is reconfigured as required to reduce proximity effects caused by proximal pattern elements in neighboring chips. To reduce variations in the imprinted profile of such an element in peripherally located chips versus centrally located chips on the substrate, portions of neighboring chips that straddle the substrate edge are imprinted nevertheless. This ensures that the edges of each entire chip imprinted on the substrate experiences the same proximity effect that is offset by the pattern defined by the reticle, regardless of whether the imprinted entire chips are located peripherally or centrally on the substrate. |
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summary | ||
description | This application claims priority to and the benefit of provisional application No. 62/433,471, filed Dec. 13, 2016, which is entirely incorporated herein by reference. The present invention generally relates to apparatus and methods for handling and storing spent nuclear fuel at nuclear power plants. At nuclear power plants having nuclear reactors, spent nuclear fuel is stored in deep reservoirs of water, called “spent fuel pools.” When a spent fuel pool reaches its spent fuel capacity limit, or when the nuclear power plant undergoes a complete removal of spent fuel from the spent fuel pool at the end of the life of the facility, the fuel is transferred into metal canisters having final closure lids that are welded closed or sealed with mechanical means at the power plants following spent fuel loading. The sealed canister is then placed into a concrete storage overpack which serves as an enclosure that provides mechanical protection, heat removal features, and radiation shielding for the inner metal canister that encloses the radioactive material. The concrete overpack, which contains the canister within which the radioactive materials are stored, is then placed in the designated secure location outside of the nuclear power plant structure, yet on owner controlled property so as to ensure proper controls and monitoring of the concrete storage overpack containing the metal canister are performed. The process of removing the spent fuel from the spent fuel storage pool, placing the spent fuel into a metal canister, and placing the metal canister into the concrete storage overpack involves the use of overhead cranes and associated lifting and rigging systems, which are thoroughly evaluated and tested prior to use to ensure that positive, safe control of the components being handled is assured during all normal and credible scenarios, including the possibility of a seismic event. The evaluations, processes, and controls employed to ensure that these components are safely handled at the nuclear power facility are subject to regulatory review and approval. Furthermore, it is important to note that the concrete storage overpack within which the metal canister is stored meets only the regulatory requirements for storage and not off-site transportation of the canisters. Regulations associated with off-site transportation require the use of a specially designed off-site transportation cask, which is quite different in design and materials from the concrete storage overpack and licensed for use by the regulatory authorities under different rules and regulations than those used to authorize concrete storage overpacks. Upon reaching the end of life for the nuclear power plant, many owners choose to completely remove the spent fuel from the spent fuel storage pool and place it into dry storage concrete overpacks, as outlined above, in an effort to minimize expenses associated with facility operation. Once the spent fuel has been relocated, the existing facility may then be demolished and removed from the owner's property. When this is accomplished, only the concrete or concrete/metal storage overpacks containing the metal canisters within which the spent nuclear fuel and radioactive materials remain. When the need to remove the metal canisters from the site arises, the metal canisters are transferred from the concrete storage overpacks to off-site transportation casks. However, since the structures containing the overhead cranes have been demolished and removed from the site, there is no installed means with which to facilitate transfer of the metal canisters from the concrete storage overpacks to the off-site transportation casks. The present disclosure provides various embodiments of a modular portable cask transfer (MPCT) facility. The MPCT facility is capable of transferring a canister containing spent nuclear fuel materials from or to a transportation cask respectively to or from a storage overpack. Telescoping legs enable movement of the transfer cask independent of the canister, which is moved using a hoist. Due to its modular configuration, the MPCT facility can be assembled, disassembled, and moved from one nuclear power plant to another. One embodiment, among others, is an MPCT facility having a plurality of parallel elongated telescoping leg assemblies. Each leg assembly has an elongated telescoping leg mounted on a movable dolly. Each telescoping leg has a longitudinal body extending from a top end to a bottom end that is mounted to the movable telescoping leg dolly. The tops ends are movable upwardly and downwardly in relation to the bottom ends. The MPCT facility includes a plurality of parallel elongated lift beams. Each of the lift beams has a longitudinal body extending from a first end to a second end. Each of the lift beams supported at the top ends of the telescoping legs and oriented in a direction that is perpendicular to the telescoping legs. The MPCT facility includes a trolley beam assembly comprising a movable elongated trolley beam mounted on a movable trolley beam dolly. The trolley beam has a longitudinal body extending between a first end and a second end. The trolley beam supported by the longitudinal bodies of the lift beams and oriented in a direction that is perpendicular to the lift beams. The MPCT facility includes a means for connecting a transfer cask to the trolley beam so that the transfer cask can be moved vertically as the top ends of the telescoping legs are moved vertically. Finally, the MPCT facility includes a hoist associated with the trolley beam. The hoist is for connecting to the canister and moving the canister vertically. The hoist can move the canister from and into the storage overpack, respectively, into and out of the transfer cask. Based upon the separate telescoping legs and the hoist, the transfer cask and the canister can be moved vertically independently of each other. Another embodiment, among others, of the MPCT facility has a gantry means for attaching to a transfer cask and for moving the transfer cask in a vertical and horizontal direction. Further, the MPCT facility has a hoist means associated with the gantry means, which is moved as the transfer cask is moved in the vertical and horizontal direction. The hoist means can connect to the canister and move the canister vertically independently of the movement of the transfer cask. The hoist means can move the canister from and into the storage overpack, respectively, into and from the transfer cask. It can also move the canister from and into the transportation cask, respectively, into and from the transfer cask. Other embodiments, methods, apparatus, features, and advantages of the present invention will be or become apparent to one with skill in the art upon examination of the following drawings and detailed description. It is intended that all such additional embodiments, methods, apparatus, features, and advantages be included within this description, be within the scope of the present disclosure, and be protected by the accompanying claims. The present disclosure provides various embodiments of a modular portable cask transfer (MPCT) facility 10. An example of an MPCT facility 10 is shown in FIGS. 1 through 5. The MPCT facility 10 can be used to transfer a canister containing spent nuclear fuel materials, such as fuel rods, from or to a transportation cask respectively to or from a storage overpack using a transfer cask as well as independent securing and control of the canister and transfer cask during the transfer operation, which is important if a seismic event occurs during the transfer operation. Due to its modular configuration, the MPCT facility 10 can be easily disassembled, moved, and reassembled so that it can be used at a plurality of nuclear power plants. Unlike a conventional gantry crane, the MPCT facility 10 has a plurality of parallel elongated telescoping leg assemblies 12 that enable it to move the transfer cask. In this embodiment, there are at least four telescoping leg assemblies. More telescoping leg assemblies may be incorporated, depending upon the needs of the facility. Each leg assembly 12 has an elongated telescoping leg 14 mounted on a movable dolly 16, preferably, a motorized, self-propelled, movable, side-shift dolly with at least four wheels and controlled by a suitable controller. The dollies 16 ride on or within tracks 17 along the top of the existing storage pad 19. Each telescoping leg 14 has a longitudinal body extending from a top end 14a to a bottom end 14b that is mounted to the movable telescoping leg dolly 16. Each longitudinal body of each telescoping leg 14 has an elongated movable outer upper part around and guided by a stationary lower inner part so that the tops ends 14a are movable upwardly and downwardly in relation to the bottom ends 14b. In this example, the upper and lower parts have square cross sections. However, the assembly cross sections can be circular, as well. Furthermore, in the preferred embodiment, a hydraulic system is designed to move the top ends 14a of the telescoping legs 14 upwardly and downwardly in relation to the bottom ends 14b. In other embodiments, a pneumatic system could be employed instead of the hydraulic system. A plurality of parallel elongated lift beams 22, two in this example, are situated at the top ends 14a of the telescoping legs 14. Each of the lift beams 22 has a longitudinal body extending from a first end 22a to a second end 22b. Each of the lift beams 22 is supported at the top ends 14a of the telescoping legs 14 and is oriented in a direction that is perpendicular to the telescoping legs 14. A trolley beam assembly 24 comprising a movable elongated trolley beam 26 is mounted on movable trolley beam dollies 28, preferably, a motorized, self-propelled, movable, side-shift dolly with at least four wheels and controlled by a suitable controller. The trolley beam 26 has a longitudinal body extending between a first end 26a and a second end 26b. The trolley beam 26 is supported by the longitudinal bodies of the lift beams 22 and oriented in a direction that is perpendicular to the lift beams 22. A plurality of lift links 32 can be used to attach the trolley beam 26 to a transfer cask 34, which can be used to remove or insert a cylindrical canister 35 made, for example, from metal, from or into a cylindrical storage overpack 37 made, for example, from concrete or concrete and metal, stored upon the storage pad 19. An example of a storage overpack 37 is described in co-pending application Ser. No. 14/160,833, filed Jan. 22, 2014, which is incorporated herein by reference in its entirety. The transfer cask 34 is made from, for example, metal and other suitable materials to provide protection from hazardous radiation, and has a plurality of lift plates 36 mounted to the transfer cask 34 to enable the transfer cask 34 to be lifted and lowered as the telescoping legs 14 are lifted and lowered. This embodiment has two lift plates 36 situated at opposing sides of the transfer cask. As shown in FIGS. 4 and 5, the transfer cask 34 has a ring 37 defining an open hole, or aperture, at its top end and two sliding metal doors 39a and 39b at its bottom end. This embodiment of the transfer cask 34 is described in further detail in co-pending application Ser. No. 14/160,833, filed Jan. 22, 2014, which is incorporated herein by reference. The sliding doors 39a and 39b, which are hydraulically or pneumatically actuated, slide outwardly sideways from the transfer cask 34. Two c-shaped channels are welded on two opposing sides at the underside of the transfer cask 34 to capture respective side edges of the doors 39a and 39b, hold them in place, and provide a sliding surface. A plurality of elongated connectors 38 connect the lift links 32 to the lift plates 36 of the transfer cask 34. The elongated connectors 38 can be, for example, slings (as shown), fixed rods, fixed links, or other apparatus, and can be made of a variety of materials, but are preferably metal. The MPCT facility 10 further includes a conventional hoist 41, preferably a chain hoist, with a hook 43 designed to connect to a canister lift adaptor 42 that is mounted to the canister 35 and designed to move the canister 35 upwardly and downwardly independent of the transfer cask 34, which is moved upwardly and downwardly via the telescoping legs 14. The hook 43 engages with the canister lift adaptor 42 and is restrained using two engagement pins 45a and 45b that essentially pin the hook 43 to the adaptor 42. The hole in the top of the transfer cask 34 is of sufficient size to enable the hook 43 associated hoist 41 to pass through the transfer cask 34. Instead of a single hoist, a dual hoist can also be used. Further, in the preferred embodiment, the chain hoist is air operated. However, in other embodiments, the hoist(s) may be an electrically operated wire rope hoist(s) or any combination of electrical or pneumatic chain or wire rope hoisting system. A plurality of elongated seismic restraint rods 44 can be employed to assist with securing the MPCT facility 10 by making the structure more rigid and therefore resistant to seismic induced lateral movements that could serve to upset the listing arrangement. In this example, there are two on each side in a crossing configuration where each restraint rod 44 connects either a top end or a bottom end of one movable outer upper part of one telescoping leg 14 to a bottom end or a top end, respectively, of another adjacent movable outer upper part of an adjacent leg 14. For further seismic stability in connection with the transfer cask 34, the MPCT facility 10 may optionally be provided with, as illustrated in FIG. 1, a ring 48 surrounding the transfer cask 34 that is supported and held in place by cables 52 that attach between the ring 48 and each of the legs 14. In essence, the ring 48 with cables 52 prevents the transfer cask 34 from swinging in the horizontal direction. Referring to FIG. 1, a transfer adaptor 54 is temporarily situated between the transfer cask 34 and the storage overpack 37 when they are engaged to assist in supporting the transfer cask 34 and to open and close the doors 39a and 39b at the underside of the transfer cask 34. The transfer adaptor 54 includes generally flat rectangular metal plate having a centralized hole through which the canister 35 is passed. The transfer adaptor 54 has indexing rails 56a and 56b on the plate along its sides to assist with positioning the transfer cask 34 and to serve as guides for the doors 39a and 39b. The transfer adaptor 54 has hydraulically or pneumatically operated cylinders 58a and 58b with corresponding rods 62a and 62b for operating (i.e., pushing and pulling) respectively the doors 39a and 39b of the transfer cask 34. The MPCT facility 10 can be used to transfer a canister 35 containing spent nuclear fuel materials, such as fuel rods, from or to a transportation cask (not shown) respectively to or from a storage overpack 37 using a transfer cask 34 as well as independently securing and controlling the canister 35 and transfer cask 34 during the transfer operation, which is important if a seismic event occurs during the transfer operation. When the canister 35 is to be moved out of the storage overpack 37 to a transportation cask, a lid 46 at the top of the overpack 37 is removed. The transfer adaptor 54 is placed on the storage overpack 37. The transfer cask 34 is then positioned over the storage overpack 37. The doors 39a and 39b of the transfer cask 34 are opened with the transfer adaptor 54. The hook 42 of the hoist 41 is lowered through the transport cask 34 and is engaged with the canister lift adaptor 42 using the engagement pins 45a and 45b. The canister 35 is lifted vertically out of the storage overpack and into the transfer cask 34. The doors 39a and 39b of the transfer cask 34 are closed with the transfer adaptor 54. The telescoping legs 14 are raised in order to raise the transfer cask 34. The transfer adaptor 54 (or another like adaptor) is placed over the transportation cask, which is open at its top for receiving the canister 35. The transfer adaptor 54 is placed on the transportation cask. The dollies 16 and 28 are used to move the transfer cask 34 with the canister 35 to a position over and then on the transfer adaptor, which is on top of the transportation cask. The doors 39a and 39b of the transfer cask 34 are opened with the transfer adaptor 54. The hoist 41 lowers the canister 35 into the transportation cask. The engagement pins 45a and 45b are retracted to disengage the hook 43 from the transfer adaptor 54. The hoist 41 raises its hook 43. The doors of the transfer cask 34 are closed with the transfer adaptor 54. The dollies 16 and 28 are used to move the transfer cask 34 away from the transportation cask. The transfer adaptor 54 is removed. When the canister 35 is to be moved into the storage overpack 37 from a transportation cask, the procedure described in the previous paragraph is simply reversed. Furthermore, the process may be used to transfer a canister from one storage overpack to another storage overpack or inspection station located on the same storage pad 19. It should be emphasized that the above-described embodiments of the present invention, particularly, any “preferred” embodiments, are merely possible examples of implementations, merely set forth for a clear understanding of the principles of the invention. Many variations and modifications may be made to the above-described embodiment(s) of the invention without departing substantially from the spirit and principles of the invention. All such modifications and variations are intended to be included herein within the scope of this disclosure. As an example of a variation, the telescoping legs 14 could have more than two nesting parts, i.e., more than an upper and lower part. As another example of a variation, the transfer cask 34, the canister 35, and/or the storage cask 37 can have other cross sectional shapes that are not circular, for example, square, rectangular, octagonal, etc., and the facility 10 of the present disclosure could be used in connection with these. |
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description | 1. Field of the Invention The present invention relates to a fuel assembly for a nuclear reactor using a coolant such as a liquid metal, and particularly to a fuel assembly which is configured to store and hold a plurality of fuel pins in a wrapper tube by using grids and liner tubes and which suppresses an unnecessary flow of the coolant in an outer circumferential side in the wrapper tube and increases the flow volume of the coolant passing through interiorly disposed ones of the fuel pins to thereby increase the core power. 2. Related Art Generally, in a nuclear reactor, a fuel assembly is supported in a reactor core while being attached to a support member. In a nuclear reactor using a coolant such as a liquid metal, an electromagnetic pump is used as a drive source to circulate the coolant around a plurality of fuel pins included in the fuel assembly supported in the reactor core. In this case, if the nuclear reactor is small-sized, the fuel assembly is configured to store the fuel pins in a wrapper tube to enable the circulation of the coolant with no need for the drive source. The wrapper tube is configured to include an entrance nozzle at a lower end thereof for introducing the coolant, and an operation handling head at an upper end thereof. The wrapper tube includes therein grids for supporting the fuel pins in the radial direction of the wrapper tube, and liner tubes inserted in the wrapper tube for fixedly holding the respective grids in the axial direction of the wrapper tube. The intervals in the radial direction of the fuel pins are kept by the grids. Meanwhile, the intervals in the axial direction of the grids are kept by a tie rod, the liner tubes, or the like (see Japanese Unexamined Patent Application Publication No. 6-174882, for example). FIGS. 23 and 24 illustrate the configuration of this type of conventional fuel assembly. In the figures, a plurality of fuel pins 101 are stored in a wrapper tube 103, with the pin intervals of the fuel pins 101 being kept by grids 102. Each of the fuel pins 101 is fixed at a lower portion thereof by a lower pin support plate 105 and at an upper portion thereof by an upper pin support plate 106. The coolant such as a liquid metal flows in from a coolant inlet 108 of an entrance nozzle 104 and flows out from a coolant outlet 109 of a handling head 107. In the thus configured fuel assembly, as illustrated in FIG. 25, each of the grids 102, which has a low pressure drop, includes a grid frame 102a provided with a multitude of ring-shaped pin support members 110. As illustrated in FIG. 26, for example, each of the pin support members 110 is provided with three dimples 110a on the inside thereof such that the circumference of the corresponding fuel pin 101 is three-point supported, for example, by the dimples 110a. FIG. 27 illustrates a deformation state in which the wrapper tube 103 is expanded by the thermal expansion. That is, the wrapper tube 103, the basic form of which is a regular hexagon as indicated by a virtual line in FIG. 27, is expanded when used due to the irradiation deformation and is deformed so as to expand toward the outer circumference thereof as indicated by a solid line. Conventionally, to cope with such deformation, liner tubes each formed by a thin hexagonal tube are provided outside a fuel bundle such that the liner tubes and the grids are alternately stacked. Thereby, the intervals in the axial direction of the grids are kept. In such a configuration, however, the flow passage area around the fuel bundle is large. Thus, the cladding temperature in a central area of the fuel bundle becomes relatively high in some cases. Therefore, there arises a need to keep the cladding temperature equal to or lower than a cladding temperature limit. As a result, the thermal efficiency is decreased. To address this issue, the inventors of the subject application have proposed a technique for reducing the cladding temperature, in which followers each having a triangular cross section are provided to reduce the flow passage area of a bundle edge sub-channel in a core heat generation unit for preventing a peripheral flow. That is, according to the technique, the liner tubes are provided with peripheral flow preventing projections to suppress the occurrence of the above-described phenomenon (see “Development of Densely. Packed and Low-Pressure-Drop Fuel Assembly for Non-Refueling Core (3),” 2004 Fall Meeting Preliminary Proceedings 307 of the Atomic Energy Society of Japan, for example). Meanwhile, in the above-described conventional configuration, the liner tubes and the grids are stacked and may be mutually misaligned in the radial direction. If the liner tubes and the grids are misaligned in the radial direction, an opening may be formed between the wrapper tube and the liner tubes to allow the coolant to flow from inside the liner tubes into the space on the wrapper tube side as a waste flow. Further, in the conventional fuel assembly, the bulging deformation occurs in the wrapper tube by the irradiation creep due to the inner pressure of the wrapper tube. It is therefore possible in the expanded portion that the flow passage area of a peripheral region around the fuel bundle is increased while the flow volume of the coolant in the central area of the fuel bundle is reduced, and thus that the cladding temperature is increased. It is also possible that the liner tubes are similarly expanded due to the inner pressure applied thereto. The present invention has been made in light of the above-described circumferences, and it is an object of the present invention to provide a fuel assembly which achieves a high thermal efficiency and a stable lifetime performance by preventing an unnecessary flow of a coolant in an outer circumferential area therein and by causing the coolant to effectively flow toward interiorly disposed fuel pins. To achieve the above object, the present invention provides a fuel assembly charged in a reactor core of a nuclear reactor using a liquid metal as a coolant. The fuel assembly includes a wrapper tube, grids, liner tubes, and a fixing device. The wrapper tube includes an entrance nozzle for introducing the coolant and an operation handling head, and stores a plurality of fuel pins. The grids are disposed in the wrapper tube to support the fuel pins in the radial direction of the wrapper tube. The liner tubes are inserted in the wrapper tube to fixedly hold the respective grids in the axial direction of the wrapper tube. The fixing device fixes the grids and the liner tubes in the radial direction of the wrapper tube. Further, in a preferable embodiment of the fuel assembly, the fixing device may include pins for fixing joining ends of the grids and the liner tubes along the radial direction of the wrapper tube. Furthermore, the fixing device may further include pin support portions, which are through holes formed on an outer circumferential side of a grid frame of each of the grids at positions corresponding to positions of engaging portions of the liner tubes, and through which the pins can be inserted in the vertical direction. The fuel assembly may further include a coolant blocking member for preventing the coolant from flowing in a gap between the inner circumference of the wrapper tube and the outer circumference of each of the liner tubes. The coolant blocking member may include contact pieces, which project from an outer circumferential side of the liner tube to come in contact with the inner surface of the wrapper tube, and which are formed of an elastic material capable of increasing the range of closure in accordance with the expansion of the wrapper tube. The coolant blocking member may be a skirt-shaped member hanging from an upper end portion of the liner tube along the outer circumferential surface of the liner tube, and may include a plurality of divided pieces divided by vertically extending grooves to individually come in contact with the inner circumferential surface of the wrapper tube. It is preferable to form the coolant blocking member from a high nickel steel. Further, the inner circumferential surface of a grid frame of each of the grids may be formed with a plurality of projections for closing gaps between outer peripherally disposed ones of the fuel pins. The projections may be formed in accordance with the pin pitch of the fuel pins. Furthermore, the fuel assembly may have a structure in which at least either one of a grid frame of each of the grids and a peripheral wall of each of the liner tubes is formed as a concave and convex wall bent toward the inner circumference thereof, and in which parts of the concave and convex wall projecting toward the inner circumference thereof close gaps between outer peripherally disposed ones of the fuel pins. An end portion of either one of the grid frame and the liner tube may be provided with closure portions for closing a space on the outer circumferential side of the parts of the concave and convex wall closing the gaps between the outer peripherally disposed ones of the fuel pins. The inner circumferential surface of each of the liner tubes may be provided with a plurality of rod members extending along the axial direction. Each of the rod members may have a substantially angular cross section and be disposed in accordance with the pin pitch of the fuel pins to close gaps between outer peripherally disposed ones of the fuel pins. Further, an upper end portion in the wrapper tube may be provided with an upper pin support plate for supporting the fuel pins, and the upper pin support plate may be pierced through by a tie rod, the upper end of which presses and holds downward the grids and the liner tubes via an elastic member. It is preferable to form the elastic member by a compression coil spring. Furthermore, a peripheral wall of each of the liner tubes may be drilled with a plurality of holes piercing through the peripheral wall to allow the coolant to flow between a space on the side of the wrapper tube and a space on the side of the fuel pins. According to the present invention, with the provision of the fixing device for fixing the end portions of the grids and the liner tubes in the radial direction, a gap can be prevented from being formed between the grids and the liner tubes by a positional misalignment in the radial direction. Therefore, the unnecessary flow of the coolant can be prevented, and the improvement of the thermal efficiency of the fuel assembly and the stabilization of the lifetime performance of the fuel assembly can be achieved. Further characteristics of the present invention will be made clearer from the following detailed description with reference to the accompanying drawings. Embodiments of a fuel assembly according to the present invention will be described below with reference to FIGS. 1 to 22. FIG. 20 is a schematic view illustrating a fourth embodiment of the present invention, and FIG. 21 is a partially enlarged cross-sectional view of FIG. 20. Further, FIG. 22 is an enlarged view of main parts of FIG. 21. The present embodiment is configured such that an upper end portion in the wrapper tube 2 is provided with the upper pin support plate 6 for supporting the fuel pins 5, and that a tie rod 21 penetrating the upper pin support plate 6 has an upper end which can be pressed down by an upper end plug 25 via an elastic member 24 such as a compression coil spring. The grids 9 and the liner tubes 8 are pressed and held downward by the elastic member 24. That is, an upper pin support ring 23 is provided at an upper end position in the wrapper tube 2, and the upper end of the tie rod 21 penetrating the upper pin support ring 23 is pressed down by the upper end plug 25 via the elastic member 24 such as a compression coil spring. Thus, the grids 9 and the liner tubes 8 are pressed and held downward by the elastic member 24 with the elastic force. According to the above-described configuration, the entirety of the components can be held by causing the upper end plug 25 of the tie rod 21 (the fuel pin 5) to press the uppermost grid 9 via the elastic member 24 in a manner such that the liner tubes 8 and the grids 9 will not be misaligned. Accordingly, even if the expansion occurs due to the heat of the fuel and the irradiation, the entirety of the components can be reliably held by causing the fuel pin 5 itself to pull the entirety of the components. At the same time, the other fuel pins 5 are allowed to freely expand. It is preferable to provide a ring having the same shape as the shape of the outer diameter of the ring element to properly apply the elastic force of the elastic member 24 to the grid 9 to thereby reliably apply the pressing force to the grid 9. With the liner tubes 8 and the grids 9 thus held with the elastic member 24 by the upper end plug 25 of one of the fuel pins 5, the fuel pins 5, the grids 9, and the liner tubes 8 can be integrally handled, and the free expansion of the other fuel pins 5 is not interrupted. The present invention is not limited to the embodiments described above, and other alterations and modifications may be made in the present invention as long as not departing from the scope of the appended claims. |
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abstract | The system and method described herein facilitate back-to-back automatic scanning of moving vehicles without have the vehicles stop in the scanning zone. The system includes a scanning zone that comprises a radiation source and a radiation source detector. The system further includes a first sensor component for automatically sensing when a first portion of the moving target has passed through the scanning zone and a second portion of the moving target is about to enter the scanning zone, wherein the first sensor component sends a signal to the automated target inspection system to initiate a scan of the second portion upon sensing that the second portion of the target is about to enter the scanning zone. Additionally, a shutter, triggered by a signal from the first sensor component, allows radiation from the radiation source to pass through the scanning zone in the direction of the radiation detector when the second portion of the moving target is passing through the scanning zone and closes off the radiation when the second portion of the moving target is no longer within the scanning zone. |
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summary | ||
description | Post-irradiation envelope measurements of some spacer designs indicate that the spacer side plates bow outwards during their incore residence time to the point where the spacer design envelope may be exceeded. The effect is most noticeable at high burnup. If severe enough and if it occurs in many assemblies, difficulties may occur when unloading or loading the core in that assemblies could get stuck and/or become difficult to extract or insert. xe2x80x9cGrowthxe2x80x9d of the spacers can also impact the seismic behavior of the assemblies in the core due to closure of the gap between adjacent assemblies. Reasons for the larger than expected growth or bow of the spacers are related to larger than expected growth or extension of the internal doublet strips that form the cells in some spacer designs. Hydrogen pick-up from the corrosion reaction causes an increase in volume of the Zircaloy material typically used to form at least part of spacers resulting in growth. The oxide that forms on the Zircaloy strip material causes tensile stresses in the underlying metal, and hence tends to lengthen the spacer strips due to creep resulting from the tensile stress. The growth or extension of the spacer strips due to these causes can be alleviated by the use of zirconium alloys with improved corrosion behavior. Measurements and calculations indicate that the sum of the effects of the corrosion related causes cannot account fully for the observed bulging of the spacer side plates during irradiation. Recent measurements during fabrication of assemblies have indicated that during loading of the rods in the assembly (or spacer), a certain amount of side plate bulging takes place. The reason for this is that in the design utilizing doublet spacer strips, the strips, when squeezed down on their springs which occurs when a rod is pushed into a cell, will increase in length because the undulations in the spacer strips are being flattened slightly. Although the effect in each spacer cell is slight, the cumulative effect of approximately 14 to 18 of the fuel rod cells causes the spacer side plates to measurably bulge outwards. An improvement in accordance with the present invention is to provide for at least some straight strips without undulations that connect the opposite sides of a spacer. Straight strips would prevent the side plates of the spacer from moving apart. At the same time, the highly desirable feature of particular spacer designs, namely the curved flow channel formed by the doublets separating adjacent fuel rods, is maintained. A typical prior art design with four channels formed by doublets and surrounding a fuel rod provides for four springs in each spacer cell. The contact point of the four springs in each cell touch on the fuel rod in a single plane. This allows the fuel rod to pivot in each cell, albeit a small amount. If the fuel rod is prone to bowing from any cause such as a small amount of initial bow, uneven heat flux or fissioning within the fuel column, buckling forces, etc., such bowing will be exacerbated by the fact that the rod can pivot in the spacer cell. A spacer or cell design that prevents pivoting of the rod in a single plane by supporting it by dimples that are spaced apart along the rod length or by supporting the rod along some length would tend to counteract rod bow and hence be desirable. In accordance with the present invention, a PWR spacer 50 is disclosed where the majority of the doublets 40 consist of one substantially flat and straight strip 42 and one strip substantially straight and with undulations 46. Each undulation 47 forms a flow channel 48 which are located between adjacent fuel rods and the majority of fuel rods is surrounded by four such flow channels. The arrangement of a fuel rod 20 within a spacer cell 52 is shown in FIG. 1. The rod is held in position by flat straight spacer strips 42 on two sides and by two flow channel springs formed in undulations which function in the same manner as the springs in the spacer described more fully in U.S. Pat. No. 4,726,926 which is hereby incorporated by reference. It is noted from FIG. 2 that the fuel rod is positioned within the cell, asymmetrically, rather than being positioned in the center of the cell. The arrangement of guide tubes, fuel rods, and spacer strips for a 17xc3x9717 pressurized water reactor fuel assembly spacer is shown in FIG. 2. Symmetry within the spacer has been maintained by having the spring strips face outward and the flat strips face inward in each spacer quadrant. A double spring strip is provided near the center of the spacer for the 17xc3x9717 array of fuel rods. The present invention can be used with virtually any other array of fuel rods. The present invention has several advantages over the prior art. Straight and essentially flat strips between nearly all the rods connect opposite sides of the spacer which prevent bulging out of the side plates during fabrication and reduces bowing-out during irradiation. This also enhances the location of the rods on a precise pitch as determined by the straight strips. This is an improvement over the prior art designs where the fuel rod is held between four springs one on each side of the spacer cell and the rod-to-rod pitch may vary because not all springs in a cell may be compressed the exact same amount. However, as contrasted to the prior art design, the fuel rods in the present invention are supported along nearly the full spacer height on two sides of the cell where they touch the straight spacer strips. This thereby eliminates pivoting of the fuel rods and reduces rod bow during irradiation. A further improvement over prior art designs is achieved by providing curved nozzles 49 with particular predetermined directions on the flow channels 48 formed by the spacer strip as more fully described below. The direction of the nozzles in accordance with the present invention is shown in FIG. 2 and in FIG. 3 showing the spacer having sides designated xe2x80x9cAxe2x80x9d, xe2x80x9cBxe2x80x9d, xe2x80x9cCxe2x80x9d, xe2x80x9cDxe2x80x9d. The nozzles of an individual doublet all point in the same direction. As indicated in FIG. 3, going from left to right across the spacer, there are four doublets with nozzles pointing towards side A as indicated by the direction of arrows, followed by four doublets pointing towards side B, four more doublets with nozzles pointing towards side A and four more doublets with nozzles pointing towards side C. This same pattern occurs for the doublets that connect side B with side D of the spacer in the figures. By pointing the nozzles in these directions, the resulting flow of coolant between the rods is generally in a diagonal direction within the array of rods which is indicated in FIG. 4. In prior art designs, coolant flow circulates around each fuel rod as shown in FIG. 5 depicted by the arrows. There is a major difference between the prior art design shown in FIG. 6A and the preset invention shown in FIG. 6B. In both designs, two coolant streams shown as arrows enter a flow channel 30 formed between four adjacent rods arranged in a square array and two streams leave each flow channel 30. In prior art designs, there are opposing flows within such flow channels which will cause significant turbulence. In the present invention shown in FIG. 6B, all the coolant within a flow channel 30 flows in the same direction, resulting in less turbulence and more effective stripping of the water film on the rod surfaces. Within an assembly, the coolant flow is in directions depicted by arrows as indicated in FIG. 7. The coolant circulates within portions of the assembly in diagonal directions, which in their entirety result in a generally circular flow pattern, one of which is indicated by the heavy outlined arrows in FIG. 7. FIG. 8 shows these flow areas within an assembly. As shown, some of the coolant is forced to leave the assembly and thereby enter an adjacent assembly, and at the same time, some coolant must enter the assembly from an adjacent assembly. Nuclear fuel assemblies are placed in the reactor core either all facing in the same direction or can be rotated by 180 degrees. In this manner, flow between adjacent assemblies is accomplished. However, an assembly rotated 90 or 270 degrees with respect to an adjacent assembly will impede the coolant flows into and out of the adjacent assemblies. The flow of coolant across the fuel rods will set up lateral forces, which if not properly balanced, could cause assembly torque and distortion and in turn could lead to fretting and possibly assembly bowing. Inspection of the resulting force vectors due to coolant flow in diagonal directions shows that the present invention results in properly balanced forces with the result that there is no net torque set up by the coolant flow. This is shown in FIG. 9. The present invention also results in an improvement in departure from nucleate boiling. Thus, in accordance with the present invention, the spacer incorporating straight spacer strips will, regardless of how the flow nozzles are directed, have the following advantageous features: maintain spacer envelope without undue bowing-out of the side plates, prevent fuel rod pivoting, and hence maintain rod straightness and provide accurate rod-to-rod pitch. In a preferred embodiment the spacer is made of a zirconium alloy with corrosion resistant properties. While the present invention has been particularly shown and described with reference to preferred embodiments thereof, it will be understood by those skilled in the art that various changes in form and details may be made therein without departing from the spirit and scope of the invention. |
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041837843 | description | DETAILED DESCRIPTION OF PREFERRED EMBODIMENTS According to the present invention, the connection of the hot gas lines, coming from the high temperature reactor, to the inlet falnges of each gas turbo apparatus is designed as a plug-in type connection. The connections of the remaining gas lines belonging to each gas turbo apparatus are located at the turbine duct and are separated from each other by peripheral seals which lie between the duct wall and the housing of the gas turbo apparatus and which are designed as plug seals. All connections between a gas turbo apparatus and the associated turbine duct are designed as connections which are detachable from the outside by remote control. For dismounting each gas turbo apparatus by means of manipulators, there are provided lifting devices which are operable from the outside through manholes. All lines which are directly connected to each gas turbo unit and which must be separated from it before dismounting of the set are detachable and, respectively, are accessible from the outside through special manholes in the prestressed concrete vessel. All bearings belonging to a gas turbo set are detachably arranged in the turbine ducts and accessible from the outside through an access duct in the prestressed concrete vessel for inspections. It is possible with a gas turbo apparatus installed according to the present invention to carry out the mounting and dismounting in a conventional manner, even though the gas turbo apparatus is arranged in a prestressed concrete vessel. The gas turbo set can be dismounted from the outside as a result of the connections between each gas turbo apparatus and the associated turbine duct which are detachable from the outside as well as with the aid of hydraulically operated manipulators acting on the lifting devices for the gas turbo unit. The anchoring of the gas turbo apparatus in the individual turbine ducts is effected, for example, by the aid of supports which transfer the basic pressure of the gas turbo apparatus to the prestressed concrete vessel. By means of hydraulic or mechanical instruments, the gas turbo apparatus can be detached from the anchoring from the outside. As far as necessary, all direct connections at the gas turbo apparatus (auxiliary systems, measuring points) are also detachable by remote control, and respectively, are arranged to be accessible from the outside of the prestressed concrete vessel. With the exception of the hot gas line connections, which are designed as plug-in type connections, all gas line connections of the coolant circuit are located at the turbine ducts, i.e., a plug-in connection is provided only at the turbine inlet, whereby the necessary cost expenditure can be reduced essentially. The gas line connections are separated from each other by peripheral seals in the form of sliding rings which are partially constructed for high pressure and which are arranged between the housing of the gas turbo apparatus and the wall of the turbine duct. The parts of the peripheral seals provided at the wall of the turbine duct are designed as plug seals. The housing of the gas turbo apparatus does not have flange connections. The metallic liners, with which the recesses serving for placement of heat exchangers and gas lines are usually coated, are flanged at the liner of the turbine duct, in those instances where recesses are in communication with the turbine duct. The bearings existing in the turbine duct for the gas turbo apparatus can be inspected and detached without requiring that the gas turbo apparatus be detached previously. The access to the bearings is assured by special access ducts in the prestressed concrete vessel. Thus, also an extensive accessibility of the gas turbo apparatus from the outside is provided. In order to facilitate the mounting and dismounting of the gas turbo apparatus, the hot gas lines are designed to be slidable in their lower part which faces toward the turbine inlet; i.e., with the aid of several lifting systems the lower parts can be telescopically inserted into the upper parts. The course of the turbine duct in the prestressed concrete vessel is subjected in the main to no limitation; the duct can lead for example horizontally or vertically through the prestressed concrete vessel. In the case of a nuclear power plant with several identical loops, which are parallel connected and symmetrically arranged in the prestressed concrete vessel, it is advantageous to provide the turbine ducts (each taking up a gas turbo unit) below the reactor cavity in a horizontal plane and to arrange them symmetrical to the vertical center axis of the prestressed concrete vessel. For example, the turbine ducts in the horizontal plane can be arranged radially (star-shaped) whereby the center of the star is lying on the vertical center axis of the prestressed concrete vessel. With an advantageous type of the nuclear reactor plant according to the invention, the turbine ducts are continued up to the vertical center axis of the prestressed concrete vessel. Each turbine duct is closed by a prestressed concrete plug. The turbine side of each gas turbo apparatus with the shaft connection for the associated generator is arranged toward the outlet of the turbine duct, while the compressor is arranged toward the center of the prestressed concrete vessel and is supported by an axial bearing, existing in the turbine duct. In a nuclear reactor plant with several loops, this bearing is located in the central area of the prestressed concrete vessel. The connection between the turbine and the generator is effected in each case by a shaft bore provided in the prestressed concrete plug. A further support for each gas turbo apparatus is provided at the prestressed concrete plug of the associated turbine duct, at which the gas turbo apparatus in addition has an axial fixed or anchor point. All supply lines belonging to a gas turbo apparatus are provided in the apparatus itself and have connections at the external bearing ends. The lead out of the supply lines from the prestressed concrete vessel is effected by a central cavity, which is provided in the area of the joinder point of the radially (star-shaped) arranged turbine ducts, and by an access duct, closed by cover plates, which opens from the bottom in the central cavity. Through these ducts the central cavity is accessible, so that the supply lines can be separated from the respective gas turbo apparatus. In order to relieve the housing of the gas turbo apparatus from the high pressure in the turbine duct, according to the invention all apparatus housings are subjected on the inside and outside to the same high pressure. Each gas turbo apparatus can be inserted as a complete unit into the respective turbine duct. In order to facilitate the mounting and dismounting, in each turbine duct there are provided rails on which in each case roller chassis can be run. Each gas turbo apparatus is equipped with at least one such roller chassis and is inserted on this chassis into the respective turbine duct. In the drawing, there is illustrated a nuclear reactor plant with three identical, symmetrically arranged and parallel connected loops as an example of an embodiment in which the gas turbo apparatus are radially (star-shaped) arranged and in which they are installed in turbine ducts, meeting in the vertical center axis of the prestressed concrete vessel. FIG. 1 shows a cylindrical prestressed concrete vessel 1, in which a graphite-moderated, helium-cooled high temperature reactor 2 is installed in a cavity 3. The coolant circuit comprises three parallel connected circuits for utilization of heat (loops), which are coupled with the reactor 2. A gas turbo apparatus 4 consisting of a turbine and a compressor, as well as a recuperator and a cooler, belong to each loop. The reactor 2 is connected with the three loops by three radial outlet flanges 5 and also by three radial inlet flanges 6. Below the floor of the reactor core, there is arranged a hot gas collecting chamber 7 for receiving heated gas leaving the core. Above the reactor core there is provided a cold gas collecting chamber 8, which receives the gas flowing back from the loops, before it is led again to the reactor core. Vertically below the high temperature reactor 2, three horizontal ducts 9 are formed in the prestressed concrete vessel 1, which are arranged radially (star-shaped) and open into a cavity 10 at their joinder point in the center of the prestressed concrete vessel 1. In each of the ducts 9, there is installed a single-shaft gas turbine 11 as well as a compressor 12, which is situated together with the turbine 11 on a common shaft 13. Each turbine 11 is coupled with a generator (not shown). Above each turbine 11 extends a vertical gas line pod 14 which is directly connected to the turbine duct 9. The three gas line pods 14 lie symmetrically on a circular line around the vertical center axis of the prestressed concrete vessel 1. In these pods three hot gas lines 15 are installed, each connecting one of the reactor outlet flanges 5 with one of the gas turbines 11. On another circular line around the axis of the pressure vessel, six vertical pods 16 are symmetrically arranged, in which the heat exchangers, i.e., recuperators and coolers (not shown), are situated. Two pods 16 are assigned to each of the three loops, one of each pair containing a recuperator and the other a cooler. Both of these pods are arranged symmetrically to one of the turbine ducts 9 and are connected in their upper part by a horizontal duct 17, serving as the gas line from the recuperator to the cooler of each loop. A further gas line 18 connects each turbine 11 with the recuperator of the same loop. The connection from the cooler to the compressor 12 of each loop is effected by a gas line 19, which is shown in the further figures (FIGS. 3a and 4a). The cooled gas is at first led through the gas transport pods 14 from the compressors 12 to the recuperators of the three loops, whereby it passes along outside the hot gas lines 15, which are designed as coaxial gas lines. Then the gas flows in each case through a horizontal connecting line 20 into those pods 16, in which the recuperators are installed. The feedback of the reheated gas from the recuperators to the reactor inlet flanges 6, is effected in each case by means of a gas line 21, which at first extends upwardly at an incline and then is formed as a coaxial (inside) line in the vertical gas line pod 14 of the respective loop. The coaxial gas lines and also all other recesses in the prestressed concrete vessel 1 are surrounded by gastight steel liners, provided with a thermal insulation and cooled by water. In the area of the coaxial gas lines in the vertical gas line pods 14, only slight loads of temperature arise at the liners, since hot gas streams are in each case surrounded by colder gas streams. In FIGS. 2a and 2b there is shown in vertical section, one of the three turbine ducts 9 in a larger scale which, as already described, discharges into a central cavity 10. The outlet of the turbine duct 9 is closed by a prestressed concrete plug 22, and the total duct is coated by a metallic liner 23. In the turbine duct 9 one of the three gas turbo apparatuses 4 is arranged in such a manner that the turbine 11 with the shaft connection for the generator lies towards the outlet of the duct 9, while the compressor 12 is situated in the center of the prestressed concrete vessel 1. In the prestressed concrete plug 22, there is provided a shaft opening or bore 24, which connects the turbine 11 via an intermediate shaft 25 with the generator. An axial supporting cap 26, mounting at the prestressed concrete plug 22, serves for supporting the gas turbo apparatus 4; the place for mounting this supporting cap on the prestressed concrete plug 22 forms an axial fixed or anchor point 27 for the gas turbo apparatus 4. The total gas turbo apparatus 4 rests on two further supports 28, one of which carries the turbine 11 and the other the compressor 12. By means of these supports, the basic pressure of the gas turbo apparatus 4 is transferred to the prestressed concrete vessel 1. FIG. 10 shows a detailed representation of the support 28 at the side of the turbine. Both supports 28 are designed in such a manner that the gas turbo apparatus 4 can be loosened from its anchoring by means of remote control. The instruments, intended for this purpose, are operated hydraulically or mechanically from the outside, and for this purpose there are provided two bores 29 in the prestressed concrete vessel 1 for each support 28. For mounting the gas turbo apparatus 4, an axial bearing 30 for the compressor 12 is arranged in the internal area of the turbine duct 9, which is, moreover, also supported by a radial bearing 31. A further bearing 32 is intended for the turbine 11. All bearings can be detached without requiring that the gas turbo apparatus must be first demounted. An access duct 33, closed by a cover plate 34, provides for the accessibility of the bearing 32 (also see FIG. 11). Both bearings 30 and 31 on the compressor 12 are passable through an access duct 35, which enters from the bottom edge of the prestressed concrete vessel in the central cavity 10, as well as through an inlet manhole 36. The access duct 35 is closed by two cover plates 37, 38 (see FIG. 5). In the central cavity 10 there exists an intermediate platform 39 for mounting and dismounting of the three gas turbo apparatuses 4 as well as the associated bearings. The inlet to the three turbine ducts 9 is closed in each case by a cover plate 40, in which there is an inlet opening 41, as well as the access to the inlet manhole 36. All access ducts are coated by metallic liners 42; likewise, the central cavity 10 has a liner coating 43. The lines necessary for the supply of the gas turbo apparatus 4, are led from the outside through the access duct 35 into the central cavity 10, from which they are transferred through the inlet manhole 36. The connections for the gas turbo apparatus 4 are arranged at the external bearing ends, so that they are accessible from the outside or can be separated. For example, in FIGS. 2a/b, 5 and 6 the supply line 44 is shown. In FIG. 6, illustrating a section through the cover plate 40, there is recognizable, furthermore, the radial attachment 45 of the housing 46 of the gas turbo apparatus 4, whereby at four points, staggered from each other by 90.degree., elements of the housing 46 and of the cover plate 40 are engaged with each other. As the FIGS. 3a/b, 4a/b and 7 illustrate, the gas line 19 coming from the cooler enters the turbine duct 9 approximately at the level of the bearings 30, 31 and also the compressor. The liner 47 of the gas line 19 is directly connected with the liner 23 of the turbine duct 9, i.e., both with this gas line, as well as with all further gas lines entering the turbine duct 9, no flange connections at the housing 46 of the gas turbo apparatus are provided. As a further example of this, there are mentioned also the gas line 18 in which the turbine exhaust gas passes to the recuperator (see FIGS. 2a/b and 11) and the gas line pod 14, in which the cold recompressed gas passes from the compressor 12 to the recuperator, coaxially to the hot gas line 15 (see FIGS. 2a/b and 8). Also here, their liners 47 are directly connected to the liner 23 of the turbine duct 9. The separation of the individual gas streams in the turbine duct 9 from each other is effected by means of peripheral seals 48, arranged between the apparatus housing 46 and the liner 23 of the turbine duct 9, which are designed as plug seals, so that they do not hinder the dismounting of the gas turbo apparatus 4. On the apparatus housing there is in each case a sealing flange, while the plugable part of the seals 48 is provided at the liner 23. In FIGS. 2a/b, the peripheral seals for the sealing of the gas streams leaving the gas lines 18 and 19 are shown. The peripheral seal 48a which faces the central cavity 10 at the gas line 19 is designed as cold gas low-pressure seal, while the other seal 48b in the area of this gas line is designed as a cold gas high-pressure seal, since it effects sealing of the compressor inlet from the compressor outlet. FIG. 2a/b shows furthermore the shaping of the turbo apparatus housing 46 in this area; in order to transfer the resulting wedge shearing forces to the liner 23, the turbo apparatus housing 45 includes a number of ribs 49. The cold gas low-pressure seal 48a is accessible through the inlet opening 41 in the cover plate 40 from the outside of the prestressed concrete vessel 1. As is shown in FIGS. 2a/b and 7, in the prestressed concrete vessel 1, there is provided a manhole 50, which opens from the bottom into the area of the compressor inlet in the turbine duct 9. Also this manhole is closed by two cover plates 51, 52 and coated by a liner 42. In the area of the turbine outlet, a manhole 53, designed in the same manner, opens into the turbine duct 9 and is also provided with cover plates (not shown) and a liner 42. Both manholes 50 and 53 represent bores for lifting and centering devices which are operated hydraulically and which serve to make possible a remote controlled mounting and dismounting of the gas turbo apparatus 4. In the turbine duct 9, there is provided in each case a support 54 for one of the lifting and centering devices, in the area of the inlet points of both manholes 50 and 53. As already described in FIG. 1 and also in FIGS. 2a/b, the hot gas line 15, which is located in the gas line pod 14, enters above the turbine 11 in the turbine duct 9. The connection of the hot gas line 15 to the turbine inlet flange is designed as plug connection 55, so that the both parts can be easily separated from each other. The hot gas line 15 is slidably designed in its lower area in such a manner that a lower part 15a can be inserted telescopically into an upper part 15b. In this manner, the hot gas line 15 can be withdrawn from the turbine duct 9, so that the total cross section of the duct is available for the mounting and dismounting of the gas turbo apparatus 4. The sliding of the hot gas line 15 is effected by means of two lifting systems 56 (FIG. 8), which work independently of each other and which are installed in two small cavities 57 above the turbine duct 9. The exact position of the lifting system 56 can be seen in FIGS. 8 and 9. The operation of the lifting system 56 is effected from the outside by means of two bores 58 provided in the prestressed concrete vessel 1. One of the driving systems 59 provided for that purpose is partially shown in FIG. 8. The turbine duct 9 is provided over the largest part of its length with guide rails 60, running parallel to each other in the bottom area, as it can be seen in FIGS. 8, 10 and 11. Four roller chassis 61, of which in each case two belong together and which form a front and a rear rolling system, are attached to the gas turbo apparatus 4. With the aid of these the gas turbo apparatus 4 can be inserted into the turbine duct 9. The front pair of the roller chassis is situated in the area of the compressor 12. |
claims | 1. An x-ray source for performing energy discrimination within an imaging system comprising:a plurality of cathode-emitting devices emitting a plurality of electrons; anda single rotating anode having a plurality of targets oriented whereupon said plurality of electrons impinge to generate a plurality of x-ray beams; anda plurality of filters simultaneously filtering said plurality of x-ray beams to generate a post-filter x-ray beam having an x-ray distribution with simultaneously a plurality of x-ray quantity versus energy peaks defined as a plurality of peaks in the number of x-rays at multiple energy levels. 2. An x-ray source as in claim 1 wherein said a plurality of cathode-emitting device comprises:a first cathode-emitting device emitting a first plurality of electrons; anda second cathode-emitting device emitting a second plurality of electrons. 3. An x-ray source as in claim 2 wherein said first cathode-emitting device emits said first plurality of electrons at a first kVp and said second cathode-emitting device emits said second plurality of electrons at a second kVp. 4. An x-ray source as in claim 1 wherein said plurality of filters comprises a stationary filter. 5. An x-ray source as in claim 1 wherein said plurality of filters comprises a rotating filter. 6. An x-ray source as in claim 1 wherein said plurality of filters comprises:a first filter filtering a first x-ray beam; anda second filter filtering a second x-ray bear. 7. An x-ray source as in claim 1 comprising:a first filter having a first energy pass range; anda second filter having a second energy pass range. 8. An x-ray source as in claim 1 wherein said plurality of targets are on different sides of said single rotating anode. 9. An x-ray source as in claim 1 wherein said plurality of targets are on different surfaces of said single rotating anode. 10. An imaging system comprising:an x-ray source comprising;a plurality of cathode-emitting devices emitting a plurality of electrons; andat least one anode having at least one target oriented relative to said plurality of cathode-emitting devices whereupon said plurality of electrons impinge to generate a plurality of x-ray beams; anda plurality of filters simultaneously filtering said plurality of x-ray beams to generate a post-filter x-ray beam having simultaneously a plurality of x-ray quantity versus energy peaks defined as a plurality of peaks in the quantity of x-rays at a plurality energy levels; andan energy differentiating detector receiving said post-filter x-ray beam and generating an x-ray signal having material energy density differentiating information. 11. An imaging system as in claim 10 comprising:a filter rotating device coupled to said plurality of filters; anda controller electrically coupled to said filter rotating device and rotating said plurality of filters. 12. An imaging system as in claim 11 wherein said controller transitions between each filter in said plurality of filters for each view within a plurality of views. 13. An imaging system as in claim 10 further comprising an x-ray detector measuring a plurality of x-ray quantity energy levels of said x-ray beam. 14. An imaging system as in claim 13 wherein said x-ray detector measures said plurality of x-ray quantity energy levels corresponding to each of said plurality of x-ray quantity peaks. 15. An imaging system as in claim 13 wherein said x-ray detector measures said plurality of x-ray quantity energy levels in at least one energy bin. |
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description | The present invention relates to the identification of deposit formations. More specifically, the present invention provides an integrated methodology for comprehensive characterization of crystals in deposits encountered in power plants on components such as in nuclear power steam generators and on nuclear fuel. During operation of a nuclear power plant, different materials are deposited upon heating surfaces of the nuclear primary system, thereby causing a change in the heating surface. In most instances, material collects upon the heating surface, thereby causing an insulating effect between the heating surfaces and the coolant of the primary system. In some instances, the material deposited upon the heating surface can cause localized corrosion and/or pitting of the surface. Operators of nuclear power systems strive to minimize the amount of deposits upon heating surfaces, thereby allowing the best possible performance from reactor systems under controlled conditions. Over time, the deposition of materials upon the heating surfaces can affect the overall economic operability of the nuclear power reactor. In order to increase the economic viability of the nuclear power station, it is desired to ascertain the exact nature of the materials deposited upon the heating surfaces as well as to determine the source of these deposits. Currently, there is no systematic, well-defined approach to the study of deposits such as nuclear steam generator deposits, or other radioactive crystalline structures in their “as found” condition in irregularities at the surface of the equipment. There is no known way to combine various electron microscopy methods in analytical electron microscopy and/or sample preparation to achieve maximum information about materials such as Chalk River Unidentified Deposits (CRUD), nuclear steam generator deposits or other radioactive deposits to determine these deposits constituents in their “as found” condition for unadulterated portions of the deposits located in irregularities at the surface of the equipment on which they are found. There is therefore a need to develop a comprehensive method to study deposits, such as nuclear steam generator deposits and CRUD, to determine the deposits crystalline structure. There is a further need for a method which allows the study of these deposits in an economical and safe manner. It is therefore an objective of the present invention to provide an integrated method for comprehensive study of deposits, such as nuclear steam generator deposits or fuel CRUD, to determine the deposits crystalline structure. It is also a further objective of the present invention to provide a method to study these deposits in an economical and safe manner. An additional objective of the invention is the adaptation and unique combination of methods of electron microscopy (EM) that comprise high resolution analytical scanning and analytical transmission utilizing multiple imaging modes, as well as selected areas of electron diffraction and energy-dispersive X-ray spectrometry. These methods can be combined to better analyze crystals found mainly in CRUD and steam generator collar deposits, in their “as found” condition, comparing select electron microscopy signals from crystal standards with the signals from the areas of interest, and those of unadulterated portions of the deposits located in irregularities at the surface of deposits under the same radioactive conditions. These methods connect morphological and analytical characterization results with a power diffraction crystal database in order to better understand crystal growth phenomenon in irregularities. According to the present application, a proposed strategy for characterization of crystals in deposits is provided mainly in nuclear power steam generators and nuclear fuel deposit CRUD flakes for a range of scales varying from 10 to 50 micron size (macrostructural analysis) to 0.1 to 10 micron (microstructural analysis) and down to 0.02 to 400 nanometers (nanostructural analysis). The present invention provides a method to analyze crystals in a deposit on a surface of a nuclear generating station heating surface that comprises extracting a sample of material from the surface of the nuclear generating station heating surface, conducting at least one of a high resolution scanning electron microscope/energy dispersive X-ray spectrometry of the sample and a scanning transmission electron microscope/selected area electron diffraction/spot and elemental mapping analysis of the sample; then conducting at least one of three-dimensional morphology, surface topography aggregation and determination of flake size/shape, phase separation and chemical composition quantification after the high resolution scanning electron microscope/energy dispersive X-ray spectrometry of the sample; then conducting at least one of an internal structure, morphology and crystal size/shape determination crystallography investigation and a chemical composition investigation after the scanning transmission electron microscope/selected area electron diffraction/spot and elemental mapping analysis of the sample. A Monte Carlo simulation of electron beam-specimen interaction is performed after the at least one of three-dimensional morphology, surface topography aggregation and determination of flake size/shape, phase separation and chemical composition quantification. Results of the high resolution scanning electron microscope/energy dispersive X-ray spectrometry of the sample, the three-dimensional morphology, surface topography aggregation and determination of flake size/shape, phase separation and chemical composition quantification and the Monte Carlo simulation are stored in a structural data base. The results of the internal structure, morphology and crystal size/shape determination, crystallography investigation and the chemical composition investigation are stored in a crystallographic data system. The method may also be performed such that the Monte Carlo simulation predicts an expected behavior of the sample under specific operating conditions. The method may also be conducted such that the step of extracting the sample of material from the surface of the nuclear generating station heating surface comprises one of collecting a CRUD sample directly on TEM grids placed on filter paper and placing a sample of standard carbon support film on top of the sample to dislodge a number of crystals from a surface of a flake of the sample of material. The method may also be performed such that the step of conducting at least one of three-dimensional morphology, surface topography aggregation and determination of flake size/shape, phase separation and chemical composition quantification after the high resolution scanning electron microscope/energy dispersive X-ray spectrometry of the sample is performed by alternating between imaging modes and changing of voltages provided from high 20 to 50 kV to a low of 0.2 to 5 kV to eliminate charging effects resulting from a radioactive field developed during analysis. The method may further be accomplished, wherein one of the three-dimensional morphology and the phase separation is determined through scanning electron microscope multimode imaging. The method may also be accomplished wherein a peak-to-background method is used during the step of conducting at least one of a high resolution scanning electron microscope/energy dispersive X-ray spectrometry of the sample to compensate for geometric effects of the sample surface. The following detailed description is provided in conjunction with the following terms used throughout. EM—electron microscope: a term used to include all types and configurations of electron microscopes. SEM—scanning electron microscope (or microscopy): an investigative device used to view minute features of a sample, the device is generally operated at accelerating voltages less than 40 kV. The scanning electron microscope relies on using a small electron probe on the sample surface to produce a signal (image) with the resolution of approximately 1 nm in a field emission scanning electron microscope; the same probe can be used to generate—excite X-rays for energy dispersive X-ray spectrometry of the same regions with approximately 1 μm spatial resolution. SE—secondary electrons: electrons emitted from the sample surface during evaluation, the electrons have an energy less than 50 eV. The electrons emitted form the SEM images as they are detected by the SEM. SEI—secondary electron imaging: an image formed from secondary electrons emitted from the sample surface that are used to determine the morphology of a sample provided to an SEM. BSE—backscattered electrons: incident electrons recoiled/reflected back from the sample by elastic collisions with the atoms. BSEI—backscattered electron imaging: an image formed from backscattered electrons providing compositional and topographic information. EBCP—electron backscattered imaging—an image formed from backscattered electrons providing compositional and topographic information. LEI—lower electron imaging: an image formed from secondary electrons using a lower secondary electron detector (SED) located below the objective lens near the sample plane. EDXS—energy dispersive X-ray spectrometry: a method of determining the sample composition by analyzing the number of X-rays of characteristic energies emitted from the sample when bombarded by an electron beam. TEM—transmission electron microscopy (or microscope): A research device generally operated at accelerating voltages >100 kV. The device has the ability to illuminate a wide area of the sample to form an image with resolutions approaching 0.1 nm or focus the probe to obtain EDXS spectra from small areas; it also has the ability to provide electron diffraction data from the same areas. BF—bright field: an imaging mode in the TEM formed by transmitted electrons. SAED—selected area electron diffraction: electron diffraction patterns obtained from a limited area of the specimen in TEM using an area-selecting aperture. EDXS—spot and elemental mapping analyses: X-ray analysis is performed by placing a small stationary probe (spot) on the sample, or by stepping the probe across the sample and obtaining an X-ray analysis at each point to construct a map of the area. In-chamber ET detector or lower SED: a detector that collects secondary electrons from the point of beam interaction (SE1) and the surrounding area (SE2). It enables a researcher to view the sample from the side, emphasizing peaks and valleys on the sample surface and show fewer effects of charging along the peaks. Through-the-lens ET detector or upper SED—a detector that collects secondary electrons (SE1) mainly from the surface of the sample. It allows one to view images from above, allowing observation into holes, crevices, irregularities or the topology/morphology of the sample. STEM—scanning transmission electron microscope: a TEM with a set of coils to scan the focused beam across the specimen as in the SEM and having secondary, backscattered and/or transmitted electron detectors to form the images of the sample. (S)TEM-scanning transmission electron microscope: an instrument capable of performing as either a TEM or STEM. FIB—focused ion-beam: a type of microscope like a SEM, but one that accelerates a focused Gallium ion beam onto the specimen instead of an electron beam. The focused beam may be used to mill away the specimen with nanometer resolution and form images from emitted secondary electrons, as in a SEM. Nano-manipulator systems: mechanical systems such as micro-tweezers that are used to pick up or manipulate submicron features of a specimen. Compositional (or Compo) mode: refers to images formed from backscattered electrons so that the images obtained correspond to specimen composition (atomic number). Topographical (or Topo) mode: refers to images formed from backscattered electrons so that the images reflect the specimen topography. SED—secondary electron detector: a detector used to collect secondary electrons to from an image. SE1—secondary electrons: secondary electrons emitted from the surface of the sample as a result of primary electron beam—sample interactions. The intensity ISE1 of the reflected primary electron beam is proportional to the coefficient of secondary emission and allows the researcher to view images from above, allowing observation into holes, crevices, irregularities or the topology/morphology of the sample. SE2—secondary electrons: secondary electrons emitted not from the point of beam interaction but from the surrounding area due to higher energy backscattered electrons. The electron intensity ISE2 is proportional to the backscattering coefficient η. The SE2 signal is a combination of secondary and backscattered electrons contributing to contrast of the image. EsB—energy and angle selective backscattered electron detector: a type of integral electron detector that uses a conductive grid to control the energies and angles of secondary and backscattered electrons that the detector collects and uses the controllable mixtures of secondary and backscattered electrons to form an image. Referring to FIG. 1, a methodology 10 for characterization of crystals of radioactive boiling deposits encountered mainly in nuclear power steam generators and nuclear fuel deposits is presented. The methodology 10 combines electron microscopy methods and methods of preparation for flakes. In the methodology 10, a first step 20 is the extraction and manipulation of samples from a source. In the present example, techniques from scanning electron microscope (SEM) and scanning transmission electron microscope ((S)TEM) examination of the samples are performed. An extraction and manipulation of the sample entails 1) adhering a sample to be tested to standard carbon SEM stubs using carbon tape. A second alternative extraction technique entails sprinkling a portion of crushed sample onto a standard carbon support film for (S)TEM analyses. The materials provided for the sample may come from scraping and/or other removal methods from the surface to be tested. The samples may be obtained from the heating surfaces of a nuclear system, such as a nuclear steam generator or a nuclear fuel rod. In the methodology 10, the use of the deposits in their “as found” state is provided. This not only allows for a rapid processing of the deposit (rapid analysis), but also lowers the variability of the results associated with working with crystals in the “as found” state. Alternatively, CRUD flakes, which are composed of a finite number of phases/crystals, are treated such that the identified phases of interest may be chosen to be analyzed. Analyzing the phases of interest by researchers can be performed continuously by repeating the procedure until a desired result (i.e. an observable result for a specified configuration) is achieved. In the methodology 10, the repeating of the procedure can be performed for a TEM analysis by collecting individual particles on carbon grids for examination in the TEM and/or alternatively (S)TEM. In the TEM, more accurate energy-dispersive X-ray spectrometry EDXS analysis are obtained as the specimens are generally thin. Consequently, probe spreading (a known error causing problem in TEM analysis) is limited and absorption of light elements is reduced. Additionally, selected area electron diffraction (SAED) patterns can be obtained for crystallographic analysis of the phases. For example, phases may be identified by determining their interplanar spacings and comparing these with tabulated values (such as in a crystallographic database) to aid in identification of the crystals present. In accordance with the present invention, a phase selection process for TEM is a position selective basis process on a CRUD flake using an exact phase selection, e.g., a focused ion beam (FIB) and/or nano-manipulators in the SEM. Samples may also be prepared by collecting CRUD crystals directly up on TEM grids, placed on filter paper, that retain the CRUD collected from the reactor fuel scraping process. Additionally, samples may be prepared by placing a sample on a surface of a standard carbon support film for(S)TEM analysis that will dislodge a number of crystals from the surface layer of the flake, creating a mirror image of the CRUD surface of interest on the carbon paper with the crystals of interest captured on it. As provided in FIG. 1, two main types of analysis are used to determine topographical, morphological and qualitative compositional information of each sample. As provided in FIG. 1, samples are identified using SEM/EDXS (Scanning Electron Microscope/energy-dispersive X-ray spectrometry) characterization methodology 100, wherein a high resolution field emission scanning electron microscope multimode examination combines a high and low secondary electron detector and backscattered electron imaging. In an exemplary embodiment of the present invention, results of such an investigation are illustrated in FIGS. 2 to 5. Additionally, FIG. 6, as presented, illustrates comprehensive topographic, morphological and qualitative compositional information regarding the sample evaluated. Three-dimensional morphology is determined through using a multimode secondary electron imaging/lower electron imaging/backscattered electron imaging. This allows for determination of a relative presence and proportion of different phases and surface topography aggregation in the sample in macro scale, micro scale and nano scale, from 1.10 to 10 micron size following the left path of analysis as described in FIG. 1, down to 100 nanometer to 0.02 size if one follows the right path of analysis as described in FIG. 1. In FIG. 6, various phases of a spinel structure found in a nuclear plant system are shown. Chemical composition of phases incorporated, as well as the composition of the spinel, can be obtained through energy-dispersive X-ray spectrometry. Results from all of the three-dimensional morphology, surface topography aggregation, flake shape/size 120 are obtained and recorded in a storage arrangement, such as a computer and compared to a structural/analytical data base 490. To identify phase separation 150 according to the average atomic number, Z, multi-mode imaging is used to provide compositional information. The image intensity resulting from the multi-mode imaging is proportional to the average atomic number. In the present invention, multimode electron imaging is used to determine the compositional information of the sample by switching back and forth between secondary electron imaging, lower electron imaging, and backscattered electron imaging modes to eliminate the effect of charging specific to the radioactive field. By utilizing this type of imaging, this imaging has the ability to separate or mix secondary electron imaging, lower electron imaging, and backscattered electron imaging signals resulting in an improved control over the signal and resolution in the images. Multimode imaging using these various signals are illustrated in FIGS. 2 to 5 for the CRUD flake BHC sample at magnifications in the range of ×1000 to ×50,000. In each of the FIGS. 2 through 5, the upper left quarter shows an upper secondary electron image (SEI). The lower left quarter shows a lower electron image (LEI) (SE1+SE2). The top right quarter shows a backscattered electron image as it provides compositional information to the image intensity proportional to the average atomic number, (Z). The lower right quarter is a backscattered electron image (labeled “topo” because the intensity reflects the sample topography). As described above, modes are switched during evaluation to eliminate charging concerns of the sample. Referring to FIG. 2, the backscattered electron image, top right quarter, reveals the location of higher atomic numbers or brighter faces on the sample surface. Lower atomic number materials have less bright surfaces for viewing. The secondary electron image, provided in the upper left quadrant, provides high resolution imaging of the surface morphology at all magnifications. The lower electron image, provided in the lower left quadrant, provides good resolution as well as compositional information or phase distinction at lower magnification. These images are less sensitive to surface charging because the backscattered electrons and secondary electrons are less dependent on surface charging. It is to be noted that as the image magnification increases, the topographical information image is increasingly different from the basic scanning electron microscope image. As is provided in scanning electron microscopy, acceleration voltages of 20 to 30 kV are used for scanning electron microscope images in order to obtain high resolution of the signal in the image. Highly accelerated voltages, such as between 20 to 30 kV, are optimal for exciting characteristic X-rays in spot and elemental mapping and analysis. For radioactive deposits or in situations where charging is a problem, low-voltage scanning electron microscopy imaging reduces unwanted charging of the sample surface, greatly improving the imaging capability although signal strength is sacrificed for heavy elements. Scanning electron microscopy is used for image captures (morphology and topography) of radioactive or heavily charged samples at low voltages (e.g. 0.5 to 5 kV) in secondary electron/backscattered electron mixing and energy and angle-selective backscattered electron detector filtering of secondary electron mode and at high voltage (20-30 kV) when obtaining chemical information in energy-dispersive X-ray spectrometry. In order to illustrate the benefits of low voltage imaging, examples of high resolution field emission scanning electron microscope images taken using a low voltage of one kilovolt in selected areas of a CRUD flake are provided as illustrated in FIGS. 7B, 7C and 7D. Referring to FIGS. 7B, 7C and 7D, these images reveal several types of crystals with unusual morphologies, including a mixture of elongated 100 to 300 nm needlelike and thin plate shaped crystals sometimes forming characteristic crystalline “flowers”, strands of twisted, long-beaded crystalline needles up to 8 nm in length and smaller sections of aggregated particles. These crystals are found in nuclear reactor heating surfaces. Observation of the samples at magnifications of ×30,000 to ×100,000 reveal structural details of the hierarchical flake structures. Compact aggregated particles exhibiting a dense packing of 100 to 300 nm diameter grains with clearly visible boundaries and ultrafine precipitates 3 to 5 nm in diameter on their surfaces are visible near the center in FIG. 7D for example. In the exemplary sample evaluated, agglomerated faceted tetrahedral and octahedral-shaped crystals show evidence of site-specific epitaxial-growth with crystalline nuclei of 20 to 50 nm in size as provided in FIG. 7E. Additionally, the thickness of thin plate shaped crystalline clusters was found to be 3 to 10 nm as provided in FIG. 7F. The images provided demonstrate a resolution that is obtained on CRUD crystals at low voltages in field emission scanning electron microscopes. Secondary electron/backscattered electron mixing and energy and angle sensitive backscattered electron detector filtering of secondary electrons allow for this resolution of the image. As provided in the methodology in FIG. 1, spot and elemental mapping analysis (EDXS imaging) is performed in step 150. Having determined the various crystal morphologies, the next step according to the present application is to obtain EDXS imaging from selected crystals to further identify the crystal's chemical composition. Selective site (or spot mode) analysis is performed under a multimode scanning electron microscope. This selective site analysis provides qualitative phase identification of the CRUD flake samples. As provided in FIG. 8A, a scanning electron microscope image of a BHC sample with locations of two spot analyses is provided as positions one (1) and two (2). Position one (1) is located on a relatively dark, micron sized particle while position two (2) is on an aggregate of smaller particles. The resulting EDXS spectra superimposed in FIG. 8B show distinct differences in the compositions of the two types of particles, with position two containing more iron as well as manganese, nickel and zinc. None of these components are present in appreciable amounts in position one. According to the present invention, for deposits presenting crystals of interest with sizes 3 μm and larger, multimode scanning electron microscopy/site-specific EDXS is an exemplary rapid analysis mode. In quantitative EDXS analysis, the accuracy of quantification (from spot spectra and elemental maps) may be questionable if the spectra are not obtained from samples that are “polished” flat over a large area with a known geometry relative to the X-ray detector. This occurs because of poorly defined measurement conditions, and the occurrence of geometric mass and absorption effects on irregular sample surface effects. To improve the interpretation and quantification of data obtained from irregular surfaces (unadulterated flake analysis surfaces) according to the present application, operating conditions for the microscope, such as the accelerating voltage, probe diameter, probe current, detector efficiency and acceptance angle, tilting angle, counting statistics, and sample related issues (such as the electrical and thermal conductivity, fluorescence induced by “hot” samples, sample stability under beam radiation, substrate material) are specified as part of the analysis. Error may occur during quantification of EDXS data obtained from irregular surfaces based on standard-less atomic number-absorption fluorescence corrections (called ZAF correction), or X-ray depth distribution (called the Phi-Rho-Z, or PRZ). This error is due to poorly defined measurement conditions and/or the occurrence of a geometric mass effect, i.e. a defined measurement condition. Additionally the occurrence of the geometric mass effect (i.e. a variation in the emitted X-ray signal due to a complex surface topography and therefore the paths the X-rays encounter in reaching the detector) and absorption effects (mainly due to severe absorption of soft OK X-rays, that result in overestimation of the concentration of heavy metals) may occur. In order to address error resolution, according to the present invention, a set of standard samples of interest for each specific deposit (e.g. Fe2O3, CuO, ZnO) has the EDXS data obtained for the samples under well defined conditions similar to deposits to be measured 170. If the sample is highly radioactive, the placement of standards for EDXS spectra will be on the grid in its immediate vicinity. This allows determination of the correction procedure that is necessary for accurate quantification of spot and elemental mapping analysis data from a particular scanning electron microscope in the same radioactive conditions. These procedures are applied to the EDXS spectra from unknown CRUD crystals to determine their compositions more quantitatively. As samples will vary in configuration, geometric effects arising from the configuration must be taken into consideration. According to the present invention, a peak-to-background method is to be used to compensate for the effects on the analysis arising from the geometry of the sample. This method specifies that the characteristic X-ray peaks and continual background radiation produced in the same region of the sample are subject to the same absorption and backscatter conditions. Measurement of the peak-to-background ratio for the elements of interest can be compared with other elements in the sample as well as established standards, to determine if significant absorption and/or fluorescence are occurring. Such measurements are particular to each microscope and detector. If significant scattering is occurring from other parts of the sample, the method may be unreliable, since the measurement depends upon measuring the local background in the same area as the characteristic X-ray lines produced. If significant scattering occurs, a Monte Carlo simulation is used to assess the size of lateral errors. For complicated geometries, the approximation is only a general indication what parameters the microscope settings should be set at. Monte Carlo simulations 500 are performed on the sample, as provided in FIGS. 9A and 9B, in order to assess the achievable lateral resolution expected during EDXS analysis from a particular material, and the effects of specific operating conditions, specimen thickness, density and chemical composition of the intensities of the emitted and absorbed X-rays as provided in FIG. 9C. These exemplary embodiments provided by the Monte Carlo method provide an indication of expected behavior of the sample. These Monte Carlo simulations are used as a guide for optimizing the microscope conditions for particular types of specimens, rather than for quantitative comparisons with unknown specimens. A structural/analytical data base 490 may be used for storing and/or comparison of the above analysis. In addition to the high resolution scanning electron microscope SEM/EDXS and site specific EDXS analysis performed, an alternative method step may be performed for analysis of samples. As provided in step 400, a scanning transmission electron microscope (S)TEM/SAED and EDXS using a conventional and high resolution imaging/electron diffraction and high spatial resolution is performed. In the methods provided in step 400, spot spectrum are used in addition to line scans, maps and spectrum imaging. As provided above, the utility of scanning electron microscope and EDXS is used for determining the morphology of the sample in question. Analytical transmission electron microscopy utilizing spot and elemental mapping analysis is highly complimentary to the scanning electron microscope methods and, in particular, enables a researcher to examine the internal structure of crystals 402, obtain EDXS analysis that are largely free from absorption and fluorescence effects/corrections 406, as well as providing electron diffraction information i.e. crystallographic information about the phases, such as their interplanar spacings and lattice type 404. Additionally, these analyses can be obtained from regions as small as 1 nm in diameter under optimal operating conditions. Thus, the spatial resolution for analytical transmission electron microscopy is an order to three orders of magnitude superior as compared to analytical field energy electrons (typically 200 kV) to pass through, or less than several hundred nanometers in thickness. Transmission electron microscopy is highly complementary to scanning electron microscopy, where the spatial resolution of spot and elemental mapping analysis is typically not better than 1000 nm. As provided in FIG. 10, a bright field (BF) transmission electron microscopy image of three CRUD particles suspended on a carbon film is illustrated. The crystals are selected from a sample and are indicated by arrows and have the same morphologies as typical submicron crystals, previously observed in a sample by scanning electron microscopy (e.g. a faceted oblique sheath (top right), a cluster of fine aggregates (top left) and irregular plates (bottom left). Consequently, these crystals are from nuclear reactor primary reactor system heating surfaces. Selected-area electron diffraction (SAED) patterns obtained from each crystal are shown adjacent to the crystals in question. The crystal in the upper-right, displays a single-crystal spot pattern, while the other two phases display ring patterns. These ring patterns indicate that they are composed of many smaller nano-crystals. To identify the phases, the d-spacings of the phases are determined from these SAED patterns and compared with d-spacings on file for various compounds in crystallographic databases. Identification of the phases of the crystals is also facilitated by simultaneously determining their compositions 406, as indicated by the three EDX spectra in FIG. 10, again located immediately adjacent to each phase, or aggregate. These EDX spectra accurately reflect the actual particle compositions because the geometric and absorption issues present in the SEM are largely mitigated in the TEM. Examination of the three EDX spectra indicates that the faceted crystal in the top-right contains a large amount of Fe, Zn and Cu, as well as Ni, Mn and minor amounts of Al and Sn. This is in contrast to the particles in the top left, which contain mainly Fe, Cu and O, and the particles in the lower left, which contain Fe, Cu and O, but also substantial amounts of Al and Si. In conclusion, FIG. 10 illustrates how the TEM procedure provides morphological 402, crystallographic 404 and compositional information 406 for submicron CRUD particles with unambiguous interpretation, different from SEM/EDXS characterization paths. Such analyses can be performed on larger crystals, but these would need to be isolated and thinned to electron transparency to do this. For the larger crystals, SEM/EDXS characterization paths may be more appropriate from an economical point of view. Spectra may be compared between different measurement types. For example, comparison of spot EDX spectra from the BHC CRUD flake acquired in the field-emission SEM with an accelerating voltage of 20 kV (FIG. 11) with one acquired in the TEM at 200 kV (FIG. 12) shows that the peak-to-background ratio in the latter case is significantly higher. This ratio is higher due to an increase in the over-voltage, or the ratio between the accelerating voltage and the voltage necessary to excite characteristic X-rays in the specimen, by factor 10, from 20 to 200 kV. As a result, absorption of soft OK X-rays is significantly lower and the ratio of the peak intensities of Fe to O, i.e., FeK/OK, decreases from 13.32 to 2.71 from the SEM to the TEM. These data indicate that the oxygen concentration in the sample should be close to 57% as obtained from the TEM, as compared to 32.9% O as estimated by the ZAF-based standardless quantification procedure in the field-emission SEM, as indicated in the accompanying tables 1 and 2 as well as FIGS. 11 and 12. Note that the spectrum and tables as well as FIG. 12 were obtained from the particles visible in the bottom-left corner of FIG. 10. Similar conclusions concerning oxygen generated X-rays would apply to the spectra in FIG. 10. As provided in FIG. 13, X-ray elemental maps acquired from the submicron sized CRUD flake particles are provided according to scanning transmission electron microscopy. This agglomerate of submicron sized particles is similar to that indicated as provided in position 2 in FIG. 8 as well as to the aggregates of similar submicron sized particles as provided in FIG. 7E. The distributions of the various elements in FIG. 12 demonstrate that there is a mixture of different phases in the agglomerate consisting of:Fe—Cu—Ni—O 1)Al—Mn—O, and 2)Ca—O enriched particles. 3) The Fe-rich phase (approximately 800 nm across) is evident extending from the top left corner and the FeK map, the Ca rich phase (400 to 600 nm in size) is present in the lower right region as evident from the CaK map and the AlMn containing phase (also approximately 800 nm in size in the maps) as is provided in the AlK and MnK maps. Due to the higher accelerating voltage and thinness of the samples, X-ray mapping of the STEM enables researchers, in the present invention, to obtain a spatial resolution approaching 1 nm, which is nearly three orders of magnitude better than that of the analytical field emission scanning electron microscope. Additionally, researchers therefore have an increased sensitivity (at least by a factor of 10) to local variations in chemical composition and lesser of distortion of soft X-rays, as mentioned previously. Referring to FIG. 13, a method step of elemental mapping in scanning transmission electron microscopy is performed to ascertain the distribution of phases in aggregates in the sample. This method step is complementary to obtaining quantitative analysis of particular areas, or phases, using the spot mode shown in the previous section. This applies to both scanning electron microscopy and scanning transmission electron microscopy at 20 kilovolts or 200 kilovolts respectively. Chemical composition quantification standards 406 may also be used to aid in analysis of (S)TEM/SAED/EDXS data. Analytical Electron Microscopy Connection with Crystal Databases Data collected through scanning electron microscopy/EDXS 100 or (S)TEM/SAED/EDXS 400 are, as provided above in the present application, connected to crystallographic material phases using one or more of the pieces of information extracted from the analysis, such as morphological information 402, crystal lattice length 404. This allows for a rapid identification of a crystal structure. The results obtained from analysis are compared to a standard for ease of identification. Results obtained are compared to the crystalline structures 410 found in the power diffraction file (PDF) crystal database from the International Center for Diffraction Data (ICDD) to determine the structures and morphologies of possible spinel, hematite and silicate crystals relating to the deposit for a number of 28 spectra are discussed below in an exemplary embodiment. To date, the latest version of PDF database allows an end user to integrate data retrieval and data analysis, thus results from SEM and other methodologies provided above are compared to the database. All entries have been put into a relational database format. In this format, all the entry fields for diffraction, crystallographic, bibliographic, and physical property data are placed in individual tables. In an exemplary embodiment, 28 energy-dispersive X-ray analysis from a radioactive deposit were examined to determine the most likely compound or compounds based on morphology, the elemental ratios and the information from the PDF Crystal database. The spot and elemental mapping and also spectra were acquired in the scanning electron microscope although several were obtained in the transmission electron microscope. The analysis results are provided in accompanying Table 1, with the result from the search of the PDF database. This table includes the spectrum identification (columns 1 and 8), the approximate compositions of the samples based on the standard-less spot and elemental mapping analysis (column 2), the identification of the sample (column 3), notable features associated with the crystals, either morphological or compositional (column 4), the likely compound type based on comparison of the compositional analysis with the PDF data (column 5), the metal/oxygen ratio obtained in the spot and elemental mapping analysis (column 6), and the iron/copper ratio in crystals containing these elements (column 7). TABLE 1Possible Compounds/Composition in Deposits According to PDF-4 File SelectionSpectrum#(lower-File (AllPossible Compound(s)*right-Kinds)Composition (at. %)SampleNotable Features(In order of possibility)Metal/OFe/Cucorner)_S006.pgt (5)Fe77Cu2Mn1O20NA-1Micron-size, needle-likeFe3O4 variation80/2077/21clusters -high Fe_S005.pgt (5)Fe75Cu2Mn1O23NA-1Rod-like submicronFe3O4 variation77/2375/22agglomerate_S001.pgt (5)Fe74Cu4O23NA-1Sub-micron agglomeratedFe3O4 variation77/2374/43particles_S002.pgt (5)Fe70Cu9Mn1O20NA-1Several-micron rodsFe3O4 variation80/2070/94_S007.pgt (5)Fe69(Cu, Mn)1O31 NA-1Same, different locationFe3O4 variation69/3169/15(Cr)trace_S004.pgt (5)Fe64Cu3Mn1O33NA-1Clusters micron-sizeFe3O4 variation67/3364/36needles/laths1Fe43Cu3O54BHCFine aggregate (few plates) Fe3O446/5443/37_S002.pgt (2)Fe87Cu3Ni1Mn1NA-1Flat particle, sub-micronFe3O4 variation92/887/38(Al, Cr, Ti, Zn)1O8_S001.pgt (2)Fe84Cu3Ni1Mn1O12NA-1Granular, sub-micronFe3O4 variation88/1284/39(Al, Zn)trace_S003.pgt (5)Fe64Cu2Zn4Mn1O29NA-1Micron-size thick plate(Zn, Mn, Fe)(Fe, Mn)2O471/2964/210(Ni)traceS001.pgt (1)Fe64Cu10Zn3Ni3(Mn, Al)1OBHCGranular, micron-sizeFe3O4 variation81/1964/1011192Fe32Cu4Zn15Al2(Mn, Ni, Sn)BHCSub-micron faceted crystal -Fe3O4, Fe2O354/4632/4121O46has Zn(1.17)3Fe22Cu5Al8Si7Mn1O57BHCLath like - has SiFe3O4, Fe2O343/5722/513(0.75)FIG. 14Same as TEM #3 aboveFe3O4, Fe2O343/5722/5(0.75)_S002.pgt (1)Fe46Cu21O33BHCGranular, sub-micron, lowFe2CuO4, Fe3O4, Fe2O357/33 (2)46/2115(Ni, Mn, Al)traceFe/CuFIG. 13Same as_S002.pgt (1) aboveHigh Cu, or low Fe/Cu ratioFe2CuO4, Fe3O4, Fe2O357/33 (2)46/21_S004.pgt (2)Fe75Cu19Zn4O2NA-1Highly faceted, micron-sizeFe2CuO4 variation98/275/1916_S004.pgt (4)Fe12Cu82O6BHCSeveral-micron covered inCu + Cu2O, Cu2O, CuO94/612/82.17granules and needles, high Cu_S006.pgt (2)Fe5Cu86O9 (Zn) traceNA-1Same, different location - highCu + Cu2O, Cu2O, CuO91/9 5/86.18Cu_S001.pgt (2)Fe20Cu58Zn2O21NA-1Large, rough particle - high Cu FeCu3Zn2O6.5, Cu2O,79/2120/5819(Ni, Mn) traceCuOFIG. 10High Cu, low FeFeCu3Zn2O6.5, Cu2O,20CuO_S001.pgt (3)Fe5Cu50Al25Zn10O11NA-1High Cu, Al, ZnCu2AlO489/11 5/50.21_S003.pgt (3)Fe7Cu51Al24Zn8O9NA-1High Cu, Al, ZnCu2AlO491/9 7/51.22_S001.pgt (4)Fe27CulAl32Zn18O23BHCSeveral-micron particle, highFe2A1O4 or FeAlZnO477/2327/123Al, Znvariation_S001.pgt (3)Fe18Cu4Al29Zn37Ni1O11NA-1High Al, ZnAl2ZnO4, Fe2CuO4 or89/1118/424FeAlZnO4 variation_S002.pgt (4)Ca33P20Fe6Cu1O41BHCHighly faceted, several micronCa3(PO4)225size_S003.pgt (4)Ca36P21Fe4Mn2O37BHCSame, different locationCa3(PO4)226(Cu)trace_S001.pgt (3)C80Fe3Cu1Al4Zn7O5NA-1High CGraphite?27_S001.pgt (3)C79Fe1Cu10Al4Zn2O5NA-1Very small particle - high CGraphite?28*List of possible compounds on following page. As is identified in Table 2, for the exemplary embodiment, most compounds appear to be some variation of Fe3O4 or similar spinel based structures with Cu, Mn, Al and Zn (and to a lesser extent occasional Ni, Cr, Ti), substituting for Fe, or one another. There are clearly Fe and Fe,Cu-based variations of this structure, e.g. spectra #1-11, as well as Cu, Al and Al, Zn variations, e.g. spectra #21-24. TABLE 2Spinel Compounds in Deposit Including Metal to Oxygen Ratio and Lattice ParametersFe3O4 orLattice Metal/OFe/CuspinelParameters.*Compoundratioratio variationa, b, c (nm)FeO10.4312Fe2O30.670.834, 0.834,0.8322(Fe0.86 Al10.14)2O30.67Fe3O40.75Yes0.8391CuO10.5118, 0.3146,0.4662Cu2O20.426CuMnO210.5898, 0.2884, 0.553Fe2CuO40.752Yes0.8216, 0.8216,0.8709Fe2ZnO40.75Yes0.8433Fe2AlO40.75Yes0.8273Al2ZnO40.75YesAl2CuO40.75Yes0.8079AlCuO20.750.2863, 0.2863,1.1314Al4Cu2O70.860.809Fe2MnO40.75YesFe2Cu0.5Zn0.5O4 Many0.754Yes0.8425variations in Cu/ZnFe2Cu0.4Zn0.6O4 Many5Yes0.8402-19variations in Cu/ZnFe2Cu0.6Zn0.4O40.754YesAl2Cu0.6Zn0.4O40.75Yes0.839Fe1.9Cu0.1Ni0.65Zn0.35O40.755Yes0.8446FeCu3Zn2O6.50.753.330.988, 0.988,0.8066Fe1.2Zn0.6Cu0.4Cr0.8O40.7519Yes0.9283Fe2.83Al18.39Cr0.78Mg7.770.920.330.8122Si0.03Zn0.07O4(Zn, Mn, Fe)(Fe, Mn)2O40.753Yes0.8458CuFeMnO40.751Yes0.84CuAlMnO40.750.5805, 0.5805, 0.828Fe2Zn0.9Mn0.1O4 0.75Yes0.8453Many variationsFe2Zn0.2Mn0.8O4 0.75Yes0.8514Many variations(Zn0.799Fe0.172Al0.029)0.8101(Fe0.02Al1.969O4) Many variations(Zn0.399Fe0.519Al0.082)0.8128(Fe0.079Al1.912O4) Many variations(Fe0.914Si0.086)(Fe0.998,0.75Yes0.8392Si0.002)2O4FeSiO20.75Fe3Al12(SiO4)31.1546FeCO30.4679, 0.4679,1.5336Ca3(PO4)20.536, 0.536, 0.7698*PDF files for all compounds above included in same order. Review of the compounds in Table 2 indicates that elements such as Cu, Fe, Mn, Al and Zn readily substitute for one another and these spinel based structures, consistent with the results above. Examination of the lattice parameters for Fe3O4 based crystals show a wide range of cubic compounds with any lattice parameter around 0.84 nm or is of the Fe3O4 phase. This indicates the ease with which these elements substitute for one another and therefore, the almost endless range of possible compositions of spinel-type faces that a composition may have. This makes unique identification based on standardless compositional analysis difficult. |
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claims | 1. A method of making a nuclear fuel duct, comprising:forming a first hollow structure having a first cross-sectional geometry by extruding or pilgering;forming a second hollow structure having a second cross-sectional geometry different from the first cross-sectional geometry by extruding or pilgering; anddisposing the first hollow structure interior to the second hollow structure to form a nuclear fuel duct;joining a side of the first hollow structure to a corner of the second hollow structure via a structural member;wherein the first hollow structure is configured to expand in at least one dimension under stress and cause at least portion of the first hollow structure to contact the second hollow structure; andwherein the second hollow structure is configured to distribute at least a portion of the stress of the first hollow structure therethrough. 2. The method of claim 1, wherein extruding comprises extruding a metal sheet, forming the metal sheet into a polygonal shape and closing the polygonal shape. 3. The method of claim 1, wherein extruding comprises extruding the first cross-sectional geometry of the first hollow structure. 4. The method of claim 1, wherein extruding comprises extruding the second cross-sectional geometry of the second hollow structure. 5. The method of claim 1, further comprising joining a side of the first hollow structure to a side of the second hollow structure. 6. The method of claim 1, further comprising compartmentalizing axially an interior of the first hollow structure. 7. A method of making a nuclear fuel duct, comprising:providing a first hollow structure having a first cross-sectional geometry configured to expand in at least one dimension under stress, the first hollow structure configured to contain a plurality of nuclear fuel elements located therein;providing a second hollow structure having a second cross-sectional geometry defining sides and corners, the second cross-sectional geometry different than the first cross-sectional geometry;placing the second hollow structure external to and around the first hollow structure to form a nuclear fuel duct such that a space is defined between the first hollow structure and the corners of the second hollow structure; andjoining a side of the first hollow structure to a corner of the second hollow structure via a structural member. 8. The method of claim 7, wherein providing the first hollow structure includes forming metal sheets into a polygonal shape and closing the polygonal shape by welding a seam, riveting, forming a seam and tack welding, forming a seam and isostatically compressing the seam, or diffusion bonding. 9. The method of claim 7, wherein providing the first hollow structure includes forming by extruding or pilgering. 10. The method of claim 9, wherein extruding comprises extruding a metal sheet, forming the metal sheet into a polygonal shape and closing the polygonal shape. 11. The method of claim 9, wherein extruding comprises extruding the first cross-sectional geometry of the first hollow structure. 12. The method of claim 7, wherein providing the second hollow structure comprises extruding the second cross-sectional geometry of the second hollow structure. 13. The method of claim 7, further comprising coupling the first hollow structure to the second hollow structure with at least one structural member. 14. A method of making a nuclear fuel duct, comprising:providing a first hollow structure having a first polygonal cross-sectional geometry configured to contain a plurality of nuclear fuel elements therein;providing a second hollow structure having a second polygonal cross-sectional geometry different than the first polygonal cross-sectional geometry; andplacing the first hollow structure internal to the second hollow structure with a portion of the first hollow structure in physical contact with the second hollow structure in order for the second hollow structure to distribute therethrough at least a portion of stress caused by expansion of the first hollow structure;wherein placing the first hollow structure internal to the second hollow structure with a portion of the first hollow structure in physical contact with the second hollow structure comprising placing a structural member connecting a side wall of the first hollow structure with a corner of the second hollow structure. 15. The method of claim 14, wherein providing a first hollow structure comprises extruding or pilgering the first hollow structure. 16. The method of claim 14, wherein providing a second hollow structure comprises extruding or pilgering the second hollow structure. 17. The method of claim 14, wherein placing the first hollow structure internal to the second hollow structure with a portion of the first hollow structure in physical contact with the second hollow structure comprising contacting a side wall of the first hollow structure with a side wall of the second hollow structure. 18. The method of claim 14, wherein providing the first hollow structure includes forming metal sheets into a polygonal shape and closing the polygonal shape along a seam. 19. The method of claim 14, further comprising axially compartmentalizing an interior of the first hollow structure. |
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051732173 | abstract | Wastes are placed in a container, and the container is closed by a lid having a port sealed by a disc. A cutting tool having a hollow drive stem pierces the disc. A grout slurry is fed through the stem so as to discharge into the container and embed the wastes in the grout when set. The wastes may be contained in a crate in the container, and may also take the form of a glove box which itself contains waste objects. |
abstract | An apparatus for transferring spent nuclear fuel in the form of a cask having a cylindrical inner shell forming a cavity configured to receive a canister containing spent nuclear fuel, an intermediate shell disposed concentrically around and spaced apart from the inner shell and an outer shell disposed concentrically around and spaced apart from the intermediate shell. A bottom flange is affixed to bottoms of each of the shells, and a bottom lid is removably affixed to the bottom flange. A top flange is affixed to tops of each of the shells, and a top lid is seated on the top flange. An annulus for air flow may be formed between the inner shell and the canister; the bottom lid may include an impact zone including impact absorbing structure; and the top flange may have integrally formed trunnions. |
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051065711 | claims | 1. A nuclear system of the type including a containment having a nuclear reactor therein, said nuclear reactor including a pressure vessel and a core in said pressure vessel, said system comprising: a heat exchanger; a pool of water surrounding said heat exchanger; means for venting said pool of water to the environment outside said containment; means for admitting a heated fluid from within said containment to said heat exchanger, whereby said heated fluid is cooled, said means including an isolation line connecting the pressure vessel with an inlet of said heat exchanger; and means for returning cooled fluid from said heat exchanger to said pressure vessel, said means for admitting further including a depressurization valve, said depressurization valve being openable to provide fluid communication from said containment outside said nuclear reactor to said heat exchanger via said pressure vessel and said isolation line. a suppression pool, said suppression pool containing a supply of water therein, there being an air-containing headspace in said suppression pool above said supply of water, a check valve for communicating said airspace with said containment, said check valve being operable to vent air in said headspace to said containment when pressure in said headspace exceeds that in said containment, a chamber, said chamber being connected to an outlet side of said heat exchanger, condensate collecting in said chamber incident cooling of said heated fluid, the means for returning cooled fluid being connected to said chamber, and a vent line connected to said chamber and extending therefrom to a vent line terminus submerged in the water supply of said suppression pool, any noncondensible gases separated from the cooled fluid being conveyed through said vent line to the suppression pool. a heat exchanger; a pool of water surrounding said heat exchanger; means for venting said pool of water to the environment outside said containment; means for admitting a heated fluid from within said containment to said heat exchanger, whereby said heated fluid is cooled, a fluid flow passing through said heat exchanger being isolated from said pool of water, said means including an isolation line connecting the pressure vessel with an inlet of said heat exchanger, and a depressurization valve for depressurizing said pressure vessel, said depressurization valve when open providing fluid communication from said environment outside said nuclear reactor to said heat exchanger via said pressure vessel and said isolation line; means for returning cooled fluid from said heat exchanger to said pressure vessel; a suppression pool in said containment, said suppression pool comprising a supply of water, there being a substantial gas space over said water supply; means for venting said nuclear reactor into said suppression pool whereby heated fluid from said nuclear reactor is cooled; a gravity pool of coolant disposed at an elevation sufficient to permit a flow of coolant from said gravity pool into said nuclear reactor pressure vessel against a predetermined pressure within said nuclear reactor pressure vessel; and means for reducing a pressure in said nuclear reactor pressure vessel to a value less than said predetermined pressure in the event of a nuclear accident wherein said means for reducing pressure in said nuclear reactor pressure vessel comprises said nuclear reactor pressure vessel venting means. a gravity pool of coolant disposed at an elevation sufficient to permit a flow of coolant into said nuclear reactor pressure vessel against a predetermined pressure within said nuclear reactor pressure vessel; means for reducing a pressure of steam in said nuclear reactor pressure vessel to a value less than said predetermined pressure in the event of a nuclear accident, said means including a depressurization valve connected to the pressure vessel, said means further including steam heat dissipating means such dissipating means including a suppression pool; a supply of water in said suppression pool, there being a headspace in said suppression pool above said water supply; a substantial amount of air in said head space; means for feeding pressurized steam from the nuclear reactor pressure vessel to a location under a surface of said supply of water, said supply of water being effective to absorb heat sufficient to reduce steam pressure below said predetermined pressure; and a check valve for communicating said headspace with said containment, said check valve being oriented to vent air in said headspace to said containment when a pressure in said headspace exceeds a pressure in said containment by a predetermined pressure differential. 2. A nuclear system according to claim 1 further comprising 3. A nuclear system of a type including a containment having a nuclear reactor therein said nuclear reactor including a pressure vessel and a core in said pressure vessel, said system comprising: 4. A nuclear system of a type including a containment having a nuclear reactor therein, said nuclear reactor including a pressure vessel and a core in said pressure vessel, said system comprising: 5. A nuclear system according to claim 4, wherein the pressure in said headspace includes a head produced by feeding steam to said location under said surface of said supply of water. |
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051195984 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS FIGS. 1 to 3 illustrate a first preferred embodiment of the present invention to construct a top slab of a nuclear containment building. As shown in FIG. 2, the side wall 20 had already been constructed in the interior of the building A. The invented technique was used to construct a top slab 22 on top of the existing wall 20 in an area defined by a double-dot broken line in FIG. 2. The first step was to construct temporary supports 26 on top of the floor slab 24 to support the top slab 22, in an arrangement which can be seen in FIG. 1 and FIG. 2. Next step was to install horizontal sleeper beams 27, which step was followed by laying of floor joists 28 perpendicular to said sleeper beams. The both beams were made of wide-flange I-beams. The slab liner 30, equipped with pre-installed anchors 31 and other ancillary components, was then laid horizontally on top of the floor joists 28. Next, using the slab liner 30 as a platform, lower reinforcing assembly 42 and lower truss structures 32 (a lower support component) were installed. The construction of the lower support structure was completed by joining said truss 32 to said liner anchor 32 with hanging bolts 41. The upper reinforcing assembly 43 was installed next. In the meantime, a ring support 36 was constructed on the top section 35a of the shielding wall 35 located in the center of the building. Next, the ring support 36 was reinforced from the inside with reinforcements 36a as shown in FIG. 2, followed by the erection of a series of support columns 37 on the circumferential top edges 20a of the wall 20, as shown in FIG. 3. It is necessary to ensure that the heights of the ring support 36 and the support columns are larger than that of the top slab. Next, upper truss structures 38 (an upper support component) were constructed horizontally and parallel to each other on top of the ring supports 36 and support columns 37. The supports on both sides of the upper truss structure 38 were provided, as shown in FIG. 3, by installing long-span trusses 39 in parallel with each other, utilizing the remaining columns 37 located on top surface of the wall 20. The long-span truss 39 was supported at suitable distances with reinforcing bundles 33, placed on top of the support column 25 standing on the floor slab 24. The upper support assemblies were constructed by joining the upper truss 38 and 39 with lower truss structure 32 by attaching hanging bolts 40 to the flange sections of the said components 38 and 39 and to the reinforcing plates 32a of the lower truss structure 32. The slab liner 30 was held in place by a large number of slab anchors 31, through hanging bolts 41 whose opposite ends were attached to the reinforcing plates 32b of the lower truss structure 32, which truss structure forming an integral lattice network for the lower support section of the top slab. After installing the lower truss structure 32, upper and lower trusses 38 and 39 and all the necessary reinforcing lattice network as shown in FIG. 1, the concrete was poured onto the surface of the slab liner 30, and into the rest of the spaces in the lower and upper reinforcing assemblies 42 and 43 to construct the top slab 22 as shown by the double-dot broken lines in FIG. 2. It should be noted that the slab liner 30 must withstand large stresses imposed by concrete weight and other construction activities; therefore, it is important to install a sufficient number of liner anchors 31 to provide proper support for the slab liner. Furthermore, if the top slab is too large to be free-standing, the temporary support components for the slab liner, support columns 26, sleeper beams 27 and floor joists 28 need not be removed so as to provide additional support for the top slab. In the construction procedure for the top slab as described above, the structural components such as sleeper beams 27, floor joists 28, slab liner 30 and liner anchors 31, the lower reinforcing assembly 42, the lower truss structure 32 and the upper support assembly 43 can all or partially be assembled on the ground and lifted to the site with a suitable lifting mechanism, such as a crane, to improve construction efficiency. By following the construction procedure for the top slab as described above, it is possible to reduce the number of support columns 25 and 26 for the top slab 22, because the support is provided by both upper truss structure 38 and the long-span truss 39, which span across the lower truss structure 32 and the wall 20. It follows that the placement restriction, due to the presence of equipment and facilities located on the floor slab 24, for the support columns 25 and 26 is also reduced. It further follows that, because of the lesser number of structural components such as support columns 25 and 26, sleeper beam 27, floor joists 28, the work is reduced of dismantling and removal of the components after the completion of the top slab construction. It should also be noted that in installing the integral fastening tools on the underside of the slab liner 30, there is less interference from the presence of structural components such as support column 25 and 26, sleeper beams 27 and floor joists 28. The load transfer characteristics of the structure shown in FIG. 1 are excellent, because the load is transferred following a logical, compatible path through the structural components supporting the overall load; from slab liner 30 to liner anchor 31, to lower truss structures 32, to upper and lower trusses 38 and 39, Furthermore, since the lower truss structure 32 is a lattice, it is easier to avoid an interference of hanging bolts 41 with the components of the lower reinforcing assembly. There are sufficient numbers of contact points with the slab liner 31, and the various components have excellent joinability. Because no part of the slab liner 30 is penetrated with structural components, said slab maintains good pressure tightness, and the pressure within the drywell (i.e. space above the floor slab 24) can be maintained. After the completion of the construction of top slab 22, ring support 36 and reinforcements 36a are removed from the top section of shield wall 35 to make way for an installation of a pressure container for nuclear generator inside the pressure shield 35. FIGS. 4 and 5 illustrate preferred embodiments for constructing a top slab of the building. The designations and symbols are the same as used in the preceding embodiment, and the explanations are omitted. It will be recalled in the preceding embodiment that the slab liner 30 was needed to be supported by temporary support columns 26 erected on the floor slab 24, because the slab liner 30 was not suspended by a more permanent lower truss structure 32. It would be more efficient to be able to eliminate such temporary support columns from the drywell space to eliminate objects interfering with the installation of equipment inside the drywell space. In this preferred embodiment, a method is presented for constructing a top slab 22, as shown by double-dot broken lines in FIG. 2, on top of the wall 20 already constructed within reactor building A. As illustrated in FIGS. 4 and 5, the first step was to install a ring liner 50, having liner anchors 51 in the interior, on the top surface 20a of the wall 20 of the building. Also, a stepping ring 52 was installed temporarily on the interior wall 35b of the shielding wall 35. Temporary support columns 53, whose height reached the tip of the ring liner 50, was erected next on the top surface of the stepping ring 52. Between the top ledges of the stepping ring 52 and the ring liner 50, were installed horizontal sleeper beams 27, upon which a pair of supporting columns 54 were erected. Floor joists 28 were then laid horizontally on top of the sleeper beams 27, and a metal slab liner 30, having pre-installed liner anchors 31, was laid on top. This was followed by an installation of a lower reinforcing assembly 42 on the top surface of the slab liner 30, and an installation of a lower truss structure 32 between the support columns 54. The lower truss structure 32 and the liner anchors 31 were joined by hanging bolts 41, and an upper reinforcing assembly 43 was then constructed on top of the lower truss structure 32. In the meantime, a ring support was constructed around the top section 35a of the shielding wall 35, and a plurality of columns was erected on the top section 20a of the wall 20. The said columns provided support for the upper truss structures 38 and the long-span truss 39 which were constructed across said columns. The upper truss members 38, 39 and the lower truss structures 32 were joined together by using hanging bolts 40 as before. When the installation of the upper truss structure 32 and the upper truss members 38 and 39 was completed, concrete was poured starting from the top surface of the slab liner 30 and proceeding on to cover the upper and lower reinforcing assemblies 42 and 43, and ultimately producing the top slab 22 of a cross section as shown by double-dot lines in FIG. 4. When the preparatory construction steps, prior to the construction of the upper truss structures 38 and the long-span truss 39, are undertaken as presented above, the load imposed by the upper and lower reinforcing assemblies 42 and 43 is temporarily support by the lower truss structure 32 through additional load support members, that is, temporary support columns 53, support columns 54, the ring liner 50, and liner anchors 51. Accordingly, it is possible to eliminate the use of the temporary supports 26 (used to support the load of the slab liner 30), which occupy drywell spaces on top of the floor slab 24 and interfere, in general, with the construction activities inside the drywell, thus permitting unconstricted planning and execution of top slab construction activities. Additionally, the need to disassemble and remove the temporary supports 26 is eliminated. It should be noted that in this preferred embodiment, the spaces inside the wall 20 also become available for construction activities other than the top slab construction. This is because the ring liner 50 and the liner anchors 51, erected on top of the wall 20, are used to provide support to the lower truss structure 32, in the absence of the upper truss structure 38 and the long-span truss 39; the present invention thereby eliminates the necessity for having additional support columns inside the walled structure. |
abstract | According to an aspect of the present disclosure, a position adjustment mechanism is provided for automatically adjusting the position of an electrode relative to a ground roller in a corona treatment apparatus. The position adjustment mechanism is capable of detecting thickened areas in a web of material that is being treated and automatically adjusting the air gap to maintain the quality of the surface treatment of the web. |
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description | 1. Field of the Invention The present invention relates to an electron beam apparatus for focusing an electron beam onto a specimen with condenser lenses and an objective lens to image electrons transmitted through the specimen and, more particularly, to an electron beam apparatus which has an electron analyzer and acts to image only electrons having an energy coincident with the characteristic absorption energy of a certain element transmitted through the specimen. 2. Description of Related Art Electron microscopes are roughly classified into two types: transmission electron microscope and scanning electron microscope. In a transmission electron microscope, an electron beam produced and accelerated from an electron gun is focused onto a specimen by condenser lenses and an objective lens. Electrons transmitted through the specimen or scattered electrons are imaged onto a fluorescent screen, photographic film, or TV camera. Sometimes, such a transmission electron microscope is fitted with an energy filter. In this instrument fitted with the energy filter, only electrons transmitted through a specimen and having characteristic absorption energy coincident with that of a certain element are imaged. Thus, information about the specimen, such as elemental distribution, can be obtained. The phenomenon where the energies of electrons are absorbed by a specimen is known as energy loss. Analysis of the energy loss is known as energy loss spectroscopy. This energy filter is often mounted behind the image. The image is detected by an image tube or the like by making use of electrons transmitted through an energy-selecting slit or aperture baffle. A certain energy loss image can be derived by appropriately setting the width and position of the slit. Although an energy loss image of a certain element in the specimen can be obtained in this way, the background level of the energy loss spectrum is high in practice. It is essential to subtract the background to extract information about the certain element from such a spectrum. Some methods are used in practice to subtract the background in this manner. These methods are described below. First, the two-window method is described by referring to FIG. 1, which is an illustration of energy loss spectrum. In this graph, the energy loss value is plotted on the horizontal axis and the electron intensity corresponding to each energy loss value on the vertical axis. In the case of this two-window method, an image (A) of energy loss value of interest is obtained. In addition, an image (B) of a lower energy loss value is obtained. Signal processing given by A−B or A/B is performed and thus information about a desired certain element is obtained. The latter calculation is performed to acquire an intensity ratio rather than background subtraction. Therefore, the ratio may be referred to as the jump ratio. Now, the three-window method is described by referring to FIG. 2. In this method, an image (A) of an energy loss value of interest is obtained. In addition, two images (B) and (C) of lower energy loss values are taken. The background of the image (A) is estimated from the two image signals (B) and (C). The value of the background is indicated by (D). Signal processing given by A−D is performed. In the above-described estimation of the background, the relation between the intensity of energy-loss spectrum and energy-loss value is derived. This depends on the actual electron energy and on the state of the specimen, i.e., contained elements and thickness of the specimen. A calculation must be performed for each individual appropriate model of the interaction between material and electron. Some interaction models have been proposed. Any one of them is used according to the element or energy to be imaged. Since such an interaction model is not directly related to the present invention, its detailed description is omitted. In any case, however, the loss energy value is switched, and plural filtered images are acquired. Computational processing is performed between the gained image signals. When images of differing losses are obtained, electrons for generating the images are introduced into an analyzer, which energy-disperses the electrons. Only those of the energy-dispersed electrons which have a certain energy value are passed through an output slit located after the analyzer. An image is obtained only from electrons having the certain energy value by an image pickup device, such as a CCD camera. Often, a sector magnetic field is used as this analyzer. The energy of the electrons passed through the output slit is varied by varying the strength of the magnetic field. An example of an electron microscope fitted with this analyzer is shown in FIG. 3, where electrons accelerated from an electron gun 1 with an accelerating voltage E are illuminated on a specimen 3 as an electron beam by an illumination lens system 2. This illumination lens system 2 includes a combination of condenser lenses and a magnetic pre-field of the objective lens. Electrons transmitted through the specimen 3 or scattered electrons are imaged onto the incident aperture baffle 6 of the analyzer 5 by an imaging lens system 4. The energies of some of the electrons are absorbed on passing through the specimen, while the energies of other electrons are absorbed according to the elements constituting the specimen. The electrons passed through an opening of the incident aperture baffle 6 enter the analyzer 5. A magnetic field is set up within the analyzer 5. The electrons incident on the analyzer 5 are deflected by the magnetic field. The angle through which the electrons are deflected differs depending on the energy. That is, the electrons are energy-dispersed by the analyzer 5. A slit baffle 7 is mounted on the exit side of the analyzer 5. The electrons passed through an opening of the slit baffle 7 are only electrons having energy E corresponding to the strength of the magnetic field in the analyzer 5. Electrons which have been deflected greatly by the analyzer 5 and have energies (E−δE) smaller than energy E are blocked out by the slit baffle 7. The electrons passed through the slit baffle 7 and having the certain energy E are imaged onto the sensitive surface of the image recording device 9, such as a CCD camera, by an imaging lens system 8. As a result, electrons having a certain energy are detected as an image by the image recording device 9. In this case, if the strength of the magnetic field forming the analyzer 5 is swept, the energy of the electron passed through the opening of the slit baffle 7 is varied by varying the strength of the magnetic field. The analyzer 5 described above forms a sector-shaped magnetic field, and the strength of the magnetic field is varied. The following configuration is also possible. The strength of the magnetic field is maintained constant. An electrically conductive tube is mounted in the electron passage within the analyzer. A constant potential is applied to the tube from a power supply 10 shown in FIG. 4 to vary the electron energy temporarily. The energy of the electron passed through the opening of the slit baffle 7 is swept. In the example of FIG. 4, the potential inside the analyzer 5 is increased, and electrons having lower energy (E−δE) are passed through the slit baffle 7. Electrons having higher energy E are blocked out by the slit baffle 7. FIG. 5 shows another example in which the energy of the electron passed through the opening of the slit baffle 7 is varied. In the configuration of this FIG. 5, the slit baffle 7 is made movable relative to the front and rear stages of electron optics. If the slit baffle 7 is moved in the direction of the arrow in the figure, electrons having different energies can be selectively passed through the opening of the slit baffle 7. In the example of FIG. 5, the opening of the slit baffle 7 is moved into the position where the electrons of the lower energy (E−δE) are imaged. On the other hand, the electrons having the energy E and imaged onto the optical axis are blocked out by the slit baffle 7. FIG. 6 shows an example in which the energy is selected without varying the conditions of the analyzer 5 and without mechanically moving the slit baffle 7. In the configuration of this FIG. 6, a deflection coil 11 is disposed between the analyzer 5 and slit baffle 7. Electrons exiting from the analyzer 5 and dispersed are deflected by the deflection coil 11. Thus, electrons having different energies can be passed through the opening of the slit baffle 7. FIG. 7 shows an example in which the energy is selected without varying the conditions of the analyzer 5, without mechanically moving the slit baffle 7, and without using a deflection coil. In the configuration of this FIG. 7, the accelerating voltage of the electron gun 1 is varied to change the energy of the electrons illuminating on the specimen. For example, the voltage with which the electrons are accelerated in the electron gun 1 is varied from E to E′ (E′=E+δE) (increased in this case). Consequently, the spectrum on the slit baffle 7 shifts. The energy loss value of the electrons passed through the slit coincides with the increment δE in the illuminating energy. That is, electrons passed through the opening of the slit baffle 7 have energy E. Electrons having the energy E and passed through the slit baffle 7 up to now come to have energy of E+δE. In consequence, the electrons are blocked off by the slit baffle 7. On the other hand, electrons having energy of E′−δE come to haveE′−δE=(E+δE)−δE=EAs a result, the electrons are bent by the analyzer 5 and pass through the opening of the slit baffle 7 on the optical axis. In this way, electrons can also be passed through an electronic slit of desired energy loss value by varying the accelerating voltage of the electron gun 1. Electron microscopes fitted with the aforementioned energy filter are disclosed in Japanese Patent Laid-Open No. 2000-268766 and Japanese Patent Laid-Open No. H11-86771. Where an image is formed by selecting electrons of a certain energy, a tube is mounted in a sector-shaped magnetic field in the beam path. A voltage is applied to the tube to vary the energy of the electrons. Moreover, a system in which a filter, such as an Ω-filter, α-filter, or γ-filter, is positioned in the electron optical system is used. As mentioned previously, four methods are conceivable to switch the loss energy. In practice, these methods are in operation. In the first method, one condition of the analyzer 5 (e.g., the strength of the sector-shaped magnetic field) is varied as shown in FIGS. 3 and 4 or a certain voltage is applied to the beam path in the analyzer and the energy of the electron is varied temporarily, thus moving the spectrum. In the second method, the exit slit baffle 7 mounted in the rear stage of the analyzer 5 shown in FIG. 5 is moved mechanically. In the third method, the deflection coil 11 is mounted between the analyzer 5 and slit baffle 7 as shown in FIG. 6. In the fourth method, the energy of the electron beam illuminated on the specimen 3 is varied by varying the accelerating voltage of the electron gun 1 as shown in FIG. 7. Of the four methods described above, the first and fourth methods have been performed widely. In the second method, the slit baffle 7 is moved mechanically and therefore, if the accuracy at which the mechanical movement is made is enhanced to a quite high level, the accuracy is unsatisfactory compared with the energy resolution. Furthermore, the reproducibility of image presents a problem. In addition, extra cost is spent for the moving mechanism. In the third method, the position of the opening of the slit fails to agree with the optical axis of the imaging lens system mounted behind the slit baffle 7 and so aberration and axial misalignment occur. In this way, the second and third methods have great problems. Consequently, the first and fourth methods are used but they still have both advantages and disadvantages. For example, in the first method, the spectrum can be moved with high reproducibility by sweeping the magnetic field in the analyzer or by maintaining the magnetic field constant and applying a potential to the tube in the beam passage within the analyzer 5. Also, axial misalignment of the electrons passed through the slit baffle 7 with respect to the imaging lens system 8 after the analyzer 5 is not produced. Furthermore, no axial misalignment occurs in the illumination lens system 2 or imaging lens system 4 before the analyzer 5 because the set conditions are not varied at all. However, in both the imaging lens system 8 after the analyzer 5 and the imaging lens system 4 before the analyzer 5, conditions (e.g., focusing) are accurately set for electrons without energy loss (zero-loss electrons) before a potential is applied to the tube in the analyzer 5. Accordingly, where the tube potential is varied, if the energy of the electrons imaged is varied by applying the tube potential, the set conditions are no longer satisfied for the electrons unless all other conditions for the lenses and deflection system are varied in relation with the tube potential. That is, defocusing occurs. In the fourth method, the accelerating voltage of the electron beam illuminated on the specimen 3 is varied. The conditions for the illumination lens system 2 in front of the specimen 3 are no longer satisfied, producing axial misalignment. However, after transmission through the specimen, desired energy-loss electrons have an actual energy coincident with the lens conditions and so the image is not defocused. Accordingly, the fourth method is generally adopted in an energy filter that selects electrons of a desired energy with the energy-selecting slit baffle 7 and brings the electrons to an image. As mentioned previously, the problem with the fourth method is that the conditions of the illuminating lens system 2 deviate. This may shift the region on the specimen 3 illuminated with the electron beam or the brightness of the illuminating electron beam may vary, degrading the accuracy of signal processing. Therefore, a method of providing feedback control has been proposed. In particular, the conditions of the illumination optical system including the illuminating lens system 2 and deflection coil 40 for axial alignment are varied according to variation of the accelerating voltage such that the region on the specimen illuminated with the beam and the brightness of the illuminating beam remain unchanged if the accelerating voltage of the electron beam is varied or increased. As described above, where the accelerating voltage of the electron beam is varied, it is necessary to vary the operating conditions of the illumination optical system 2, because the strengths of the lenses and the strength of the deflection coil are in proportion to the square root of the relativistic energy of each lens. More specifically, let E be the energy prior to increase of the accelerating voltage. Let E* be the relativistic energy. Let E′ (=E+δE) be the energy of the electron beam after the accelerating voltage is increased. Let E′* be the relativistic value of this energy E′. There is the following relation among the energy not yet increased, the current I flowing into the lenses and deflection coil, and the current I′ flowing into them after the increase: I ′ I = E ′ * E * For these reasons, where the accelerating voltage of the electron beam is varied, the operating conditions of the lenses and deflection coil of the illumination optical system are controlled by feedback to prevent positional deviation of the electron beam on the specimen 3 and brightness variations. In the normal transmission electron microscope, the specimen is placed within the magnetic field of the objective lens 20. The magnetic field before the specimen acts as an illumination lens. The magnetic field after the specimen acts as an imaging lens. This means that correct operation cannot be expected unless feedback to the illumination optical system located ahead of the specimen is also applied to the objective lens 20. It is impossible, however, in practice to control the illuminating action of the objective lens 20 and the imaging action separately. In spite of this, if the strength of the magnetic field of the objective lens 20 ahead of the specimen 3 is controlled by feedback according to variation of the accelerating voltage, the imaging action of the magnetic field of the objective lens 20 behind the specimen 3 is adversely affected. This defocuses the image. As a result, the purpose cannot be achieved with the feedback to the illuminating lens system. It is an object of the present invention to provide an electron beam apparatus having an electron analyzer capable of controlling the illuminating lens system by feedback without adversely affecting the imaging action even if a specimen is placed within the magnetic field of the objective lens 20. An electron beam apparatus having an electron analyzer according to the present invention has: an illumination optical system consisting of lenses and deflection means for illuminating electrons at a specimen, the electrons being produced and accelerated from an electron gun; an imaging optical system for imaging electrons transmitted through the specimen positioned within the magnetic field of an objective lens 20; and the electron analyzer having a detection system for detecting the imaged electrons and energy selection means for energy-dispersing the detected electrons and selecting electrons having a certain energy. The accelerating voltage of the electron gun is varied to shift the energy of the detected electrons. Signals supplied to the lenses and deflection means of the illumination optical system are corrected using amounts of correction each obtained by multiplying an energy shift value corresponding to a variation in the accelerating voltage by a corrective coefficient. As a result, where the accelerating voltage of the electron gun is varied to cause an energy shift, the operating conditions of the illumination lens system are prevented from deviating; otherwise, the region on the specimen illuminated with the electron beam would be shifted or the illumination brightness of the beam would vary. Other objects and features of the invention will appear in the course of the description thereof, which follows. Embodiments of the present invention are hereinafter described in detail with reference to the accompanying drawings. FIG. 8 shows a transmission electron microscope according to the present invention. This microscope has an electron gun 21 producing and accelerating an electron beam. The beam is condensed by an illumination optical system 22 and illuminated at a specimen 23. The illumination optical system 22 includes plural condenser lenses 24 and deflection coils 25 for axial correction. The specimen 23 is held in a specimen holder 26 mounted to a specimen stage (not shown). An imaging optical system 27 for imaging a TEM image is mounted behind the specimen holder 26 and fitted with plural lenses and plural deflection coils. An electron energy analyzer 28 is mounted behind the imaging optical system 27. In the present embodiment, a sector-shaped magnetic field is used as the analyzer. Incident electrons are dispersed within the analyzer according to their energies. An Ω-filter, α-filter, or γ-filter can be used as the energy analyzer 28. An energy-selecting slit baffle 29 for passing only electrons having a selected energy is positioned behind the analyzer 28. A pre-slit imaging optical system 30 for imaging energy-dispersed electrons emerging from the analyzer 28 onto the slit baffle 29 is mounted ahead of the slit baffle 29. A post-slit imaging optical system 32 for imaging electrons of a certain energy passed through the opening of the slit baffle 29 onto a detector 31, such as a CCD camera, is mounted behind the slit baffle 29. The output signal from the detector 31 is supplied to a display device or signal processor 40 for obtaining an energy-loss spectrum. An accelerating-voltage control module 33 is connected with the electron gun 21. The accelerating voltage of the electron gun 21 is controllably applied by this control module 33. An optics control module 34 is connected with the lenses and deflection coils contained in the illumination optical system 22 and imaging optical system 27. The currents and voltages applied to these lenses and deflection coils are controlled by this optics control module 34. The energy analyzer 28, pre-slit imaging optical system 30, and post-slit imaging optical system 32 are controlled by a filter optics control module 35. The accelerating voltage control module 33, optics control module 34, and filter optics control module 35 are under control of a CPU 36. When an accelerating voltage is specified from the CPU 36, the accelerating voltage control module 33 sets the accelerating voltage to be applied to the electron gun 21 to the specified value. The CPU 36 also controls the optics control module 34 to control the currents supplied from the optics control module 34 to the lenses and deflection coils contained in the illumination optical system 22 and in the imaging optical system 27. Furthermore, the CPU 36 controls the filter optics control module 35 such that an electron image of a certain energy is imaged onto the sensitive surface of the detector 31. In addition, the CPU 36 controls an energy shift control module 37, which, in turn, controls the accelerating-voltage control module 33 to vary the value of the accelerating voltage, thus shifting the energy of the electron beam. An energy shift feedback control module 38 under control of the energy shift control module 37 controls the optics control module 34 to vary the amounts of current supplied to the lenses and deflection coils according to the accelerating voltage. The operation of the instrument constructed in this way is described below. The electron beam produced and accelerated from the electron gun 21 is condensed by the illumination optical system 22 and illuminated at the specimen 23. At this time, the accelerating voltage of the electron gun 21 is set from the CPU 36 via the accelerating-voltage control module 33. The illumination conditions under which the electron beam is illuminated at the specimen 23 are controlled by the optics control module 34. Electrons transmitted through the specimen 23 are imaged by the imaging optical system 27. The magnification of the image and the conditions under which the electrons enter the energy analyzer 28 are controlled by the optics control module 34. The analyzer 28 spectrally resolves the incident electrons and guides them to the pre-slit imaging optical system 30 of the filter imaging optical system. This pre-slit imaging optical system 30 guides the spectrally resolved energy spectrum to the energy-selecting slit baffle 29, and acts to enlarge the spectrum and correct aberration or distortion. Note that the pre-slit imaging optical system 30 is not essential. A transmission electron microscope fitted with an energy filter that does not have this optical system also exists. The slit baffle 29 acting as a filter causes only electrons having an appropriate energy width centered at a selected energy value to pass through the opening of the slit. The electrons passed through the slit baffle 29 enter the post-slit imaging optical system 32. This optical system 32 enlarges the incident electrons and images a projection image of the specimen 23 onto the sensitive surface of the rear-stage detector 31 operating as an image tube, such as a CCD camera. As a result, an energy-loss image of the specimen owing to electrons having the appropriate energy width centered at the selected energy is displayed on the display device 40, such as a CRT or liquid-crystal panel connected with the detector 31. In the present embodiment, when an energy-loss image owing to electrons of different energies is acquired, the accelerating voltage of the electron gun 21 is varied. The flow of processing for acquiring an image while shifting the electron energy is described below. First, the operator manipulates the mouse or keyboard connected with the CPU 36 to perform an operation for shifting the energy. Then, the CPU 36 gives instructions for setting the energy shift to the energy shift control module 37. This control module 37 issues instructions for shifting the accelerating voltage to the specified value to the accelerating voltage control module 33 and transfers information about the energy shift to the energy shift feedback control module 38. This feedback control module 38 calculates a feedback value based on the information about the energy shift and on feedback conditions defined separately and supplies corrective information about the lenses and deflection coils to the TEM optics control module 34. The illumination optical system of the TEM undergoes feedback control based on the instructions for the energy shift in this way. The amount of feedback is found from the above-described equation, i.e., I ′ I = E ′ * E ′ ( 1 ) As mentioned previously, in this equation, E is the energy prior to shifting of the accelerating voltage, E* is the relativistically modified value of the energy, E′ (=E+δE) is the energy of the electron beam after the accelerating voltage has been increased, and E′* is the relativistically modified value of this energy E′. From the above equation, the amount of correction δI of the illumination optical system is found to be equal to I−I′ (δI=I−I′). This amount of correction is applied to the lenses 24 and deflection coils 25 of the illumination optical system 22. The amount by which the electron beam is deflected by the lens strength and deflection coils is corrected according to the amount of shift of the accelerating voltage. This correction is not applied to the objective lens 20. Since the lens strength of the objective lens 20 is not feedback-controlled, it cannot be said that the illumination optical system 22 is completely corrected by a shift of the accelerating voltage. That is, this is due to the fact that the magnetic field of the objective lens 20 produced ahead of the specimen is not corrected in a corresponding manner to the shift of the accelerating voltage. The instrument is so designed that the correction for the objective lens 20 is assigned to other illumination lenses 24 and deflection coils 25. Specifically, the amount of correction for the illumination lenses 24 relative to the energy shift value is measured in advance and stored in the memory within the energy shift feedback control module 38. Similarly, the value of the deflection coils 25 relative to the energy shift value is measured in advance and stored in the memory within the energy shift feedback control module 38. Accordingly, the energy shift feedback control module 38 finds the amount of lens correction and the amount of deflection coil correction by performing the following calculations:[Amount of lens correction]=[Corrective Coefficient of Lens]×[Energy shift value][Amount of deflection coil correction]=[Corrective Coefficient of deflection coil]×[Energy shift value] (2) The corrective coefficients of the above equations can be calibrated. In the aforementioned feedback control, one condenser lens 24 in the illumination optical system 22 is used as the lens for correcting the lens strength. Alternatively, a combination of plural condenser lenses or all condenser lenses may be corrected. Similarly, in the above feedback control, one deflection coil 25 in the illumination optical system 22 is used. Alternatively, a combination of plural deflection coils or all deflection coils may be corrected. Furthermore, an additional lens or deflection coil may be mounted for correction. The calibration is next described. To facilitate the understanding, it is assumed here that one of the corrective condenser lenses 24 is corrected in terms of strength and that one of the corrective deflection coils 25 is corrected in terms of deflecting field. The aforementioned corrective coefficients depend on the operating conditions of the illumination optical system and on the energy shift value to be achieved. Accordingly, if necessary, the corrective coefficients need to be reset. That is, a calibration is necessary. The procedure of this calibration is as follows. First, desired illumination conditions (such as the illuminate position of the electron beam and the illumination size) are adjusted at some energy shift value δE1 (e.g., 0 eV meaning no energy shift). The value I1 of the current through the corrective condenser lens 24 and the values IX1 and IY1 of the current through the corrective deflection coil 25 which are taken at this time are stored in the memory within the feedback control module 38. Then, the energy shift value is set to δE2. The values of the currents supplied to the corrective condenser lens 24 and corrective deflection coil 25 are adjusted to produce the same illumination conditions as the illumination conditions (such as the illuminate position of the electron beam and the illumination size) produced when the energy shift value was δE1. The value I2 of the current through the corrective condenser lens 24 and the values IX2 and IY2 of the current through the corrective deflection coil 25 which are produced at this time are stored in the memory within the feedback control module 38. Corrective coefficients KI, KDx, and KDy are calculated based on the values found in this way, i.e., the value of the current through the corrective condenser lens 24, the values I1, I2 of the current through the corrective deflection coil 25, and the values IX1, IY1, IX2, and IY2 of the current through the corrective deflection coil 25. The calculations are performed by the energy shift feedback control module 38 based on the following equations:KI=(I2−I1)/(δE2−δE1)KDx=(IX2−IX1)/(δE2−δE1)KDy=(IY2−IY1)/(δE2−δE1) The corrective coefficients are found by the procedure described above. The corrective values δI, δIX, and δIY for the energy shift value δE are calculated using the following equations.δI=KI×δEδIX=KDx×δEδIY=KDy×δE The corrective values found by the above-described calculations are supplied to the TEM optics control module 34. In this module, the current value to the condenser lens 24 is corrected using the amount of correction δI. Also, the amount of deflection to the deflection coil 25 in the X-direction is corrected using the amount of correction δIX. The amount of correction in the Y-direction is corrected using the amount of correction δIY. As a result, even where the accelerating voltage of the electron gun 21 is varied to thereby vary the selected electron energy, the position and brightness of the electron beam illuminating the specimen 23 are prevented from being affected. While one embodiment of the present invention has been described so far in connection with FIG. 8, the invention is not limited to the structure shown in FIG. 8, but rather other modifications are possible. For example, in the embodiment of FIG. 8, each corrective coefficient is a linear function, i.e., a straight line. If higher-order functions are adopted, more accurate corrections may be made. In the embodiment described so far, lenses and deflection coils that are corrected are of the existing constructions. New lens and deflection coil may be provided and used to correct the electron beam that is made to illuminate the specimen when an energy shift is caused. Furthermore, in the above embodiment, the accelerating voltage of the electron gun is varied to cause an energy shift. Thus, the operating conditions of the illumination optical system are corrected according to above-mentioned Equation (1). Additionally, the magnetic field component of the objective lens 20 produced ahead of the specimen is corrected by the lens and deflection coil located ahead of the objective lens 20 according to above-mentioned Equation (2). However, practically sufficient advantages can be obtained by correcting the illumination lenses and deflection coils according to above-mentioned equation (2) without correcting the illumination optical system according to above-mentioned equation (1). In this case, it is preferable to calibrate the corrective coefficients correctly. It is also to be understood that the invention can be applied to every instrument that obtains elemental information contained in the specimen by energy-loss electron spectroscopy. In the embodiment described already in detail in connection with FIG. 8, electrons are energy-dispersed by a single sector-type magnet. The invention can also be applied to an instrument in which electrons are dispersed by plural magnets and to an instrument using any type of analyzer in which electrons are energy-dispersed by an Ω-filter, α-filter, γ-filter, or the like. Of course, as described previously, the invention can be applied to the magnetic field type for dispersing electrons. Besides, the invention can be applied to an instrument using electrostatic deflection coils or electrostatic mirrors and to an instrument using such electrostatic deflection coils or mirrors in combination with magnets. As described so far, an electron beam apparatus having an electron analyzer according to the present invention has an illumination optical system consisting of lenses and deflecting means for illuminating electrons at a specimen, the electrons being produced and accelerated from an electron gun, an imaging optical system for imaging electrons transmitted through the specimen positioned inside the magnetic field of the objective lens, a detection system for detecting electrons, and energy selection means for energy-dispersing electrons and selecting electrons having a certain energy. This apparatus is characterized in that the accelerating voltage of the electron gun is varied to shift the energy of electrons and that signals supplied to the lenses and deflection means of the illumination optical system are corrected using amounts of correction each obtained by multiplying an energy shift value corresponding to a variation in the accelerating voltage by a corrective coefficient. As a result, where an energy shift is caused by varying the accelerating voltage of the electron gun, shift of the illuminated region on the specimen and variations in the illumination brightness of the electron beam are prevented if the operating conditions of the illumination lens system deviate. Furthermore, the strength of the magnetic field of the objective lens formed ahead of the specimen can be corrected by calibrating the corrective current values and appropriately adjusting the values of the currents flowing through the lenses and deflection coils in the illumination optical system. |
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044619545 | summary | FIELD OF THE INVENTION The present invention relates in general to ion-processing techniques such as machining (material removal), etching, implantation, plating and cleaning by energetic ions. More particularly, the invention is concerned with an improved method of and apparatus for processing a workpiece with a beam of energetic ions whereby the mean free path of energetic ions with which the workpiece is bombarded is regulated. BACKGROUND OF THE INVENTION Conventional ion-processing apparatus is characterized by massive and complicated equipments including separately an ion chamber, an acceleration chamber, a focusing section and a workpiece mounting chamber. Since the chambers are massive, they cannot easily be evacuated. In addition, difficulties have been encountered in assuring an adequate localization of ions impinging on a desired area and a uniformity of the density of the impinging ions on a localized area. Also, the processing rate has generally been limited to an unsatisfactory level. Thus, the conventional ion-processing art has left much to be desired not only as regards equipment but also as regard to the processing precision and efficiency. OBJECTS OF THE INVENTION It is, accordingly, an important object of the present invention to provide an improved ion-processing method capable of achieving the machining, etching, implantation, plating and cleaning of a workpiece surface with an energetic beam of ions with increased efficiency and precision. A further important object of the invention is to provide an improved ion-processing apparatus which is adapted to carry out the improved method and is relatively simple in structure and organization and relatively compact. SUMMARY OF THE INVENTION In accordance with the present invention there is provided, in a first aspect thereof, a method of processing a workpiece with a beam of energetic ions, which method comprises the steps of: positioning a slender tubular member to bring its open end into spaced juxtaposition with the workpiece across a small gap of a size ranging between 10 and 1000 .mu.m in an evacuated space; supplying the tubular member with an ionizable material in the form of gas or vapor (hereinafter: ionizable gas) for feeding it into the small gap through the open end; energizing the supplied gas to form ions thereof and applying an electrical potential to the formed ions to propel them in a beam across the small gap to impinge onto a limited area of the surface of the workpiece juxtaposed with the open end of the slender tubular member; and maintaining the pressure within the small gap in excess of the pressure of the space surrounding the small gap. The small gap defined between the open end of the slender tubular member and the workpiece is preferably of a size not greater than 100 .mu.m and more preferably not in excess of 50 .mu.m. In addition, the pressures of the space and within the small gap are maintained preferably in the range between 10.sup.-6 and 10.sup.-4 Torr and in the range of 10.sup.-4 and 10.sup.-1 Torr, respectively. The pressure within the small gap is maintained desirably at least by one order of magnitude in Torr greater than the pressure of the space. The gas forming ions may be argon, nitrogen, hydrogen and/or oxygen. The gas may also be of a substance selected from the group which consists of polyhalogenated hydrocarbons containing fluorine and chlorine (marketed as "Freons"), fluorides and chlorides. These gases are chemically reactive with the workpiece material and thus can provided the action of chemical reaction in addition to the mechanical action of energetic ions impinging upon the workpiece. The invention also provides, in a second aspect thereof, an apparatus for processing a workpiece with a beam of ions, which apparatus comprises: a slender tubular member having an open end and communicating with an inlet conduit; means for positioning the slender tubular member to bring the open end into spaced juxtaposition with the workpiece across a small gap of a size range between 10 and 1000 .mu.m in an evacuated space; means for supplying an ionizable gas into the slender tubular member through the inlet conduit and feeding the gas into the small gap through the open end; means for energizing the supplied gas to form ions of the thereof and applying an electrical potential to the formed ions to propel them in a beam across the small gap to impinge on a limited area of the surface of the workpiece juxtaposed with the open end of the slender tubular member; and means for maintaining the pressure within the small gap in excess of the pressure of the space surrounding the small gap. Specifically, the last-mentioned means includes a receptacle for receiving the workpiece and at least a portion of the open end of the slender tubular member and vacuum pump means for maintaining the pressure of the space within the receptacle in range between 10.sup.-6 and 10.sup.-4 Torr. The supply means includes valve means constituting a portion of the pressure-maintaining means and arranged between the inlet conduit and a source of the gas. The valve means is adjustable to maintain the pressure within the small gap in a range between 10.sup.-4 and 10.sup.-1 Torr. Preferably, means is provided which is responsive to a change in the pressure within the small gap for controlling at least one of the vacuum pump means and the valve means. In addition, means is provided which is responsive to a change in the gap size for relatively displacing the slender tubular member and the workpiece so as to maintain the gap size substantially at a predetermined value. Means is also provided for relatively displacing the workpiece and the slender tubular member in a plane substantially orthogonal to the longitudinal axis of the slender tubular member along a prescribed path to successively process the workpiece surface in a scanning manner over a preselected area thereof. It is desirable that the tubular member be composed at least in part of an electrically conductive material to constitute one of a pair of electrodes with respect to the workpiece constituting the other electrode for establishing the said electrical potential therebetween. Preferably, the slender tubular member is composed at least at a portion of the open end of such electrically conductive material. The tubular member may have an inner diameter ranging between 0.1 and 5 mm. Preferably, means is further provided for applying a magnetic field to the beam of energetic ions in the region of the small gap to controlledly facilitate dispersion of the ions in the beam impinging on the workpiece. The magnetic field should preferably be of a flux density not less than 500 Gauss. |
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abstract | Provided is a plasma diagnosis system using multiple-reciprocating-pass Thompson scattering. The plasma diagnosis system includes: a laser which supplies laser pulses; an optical system configured to make the laser pulse multiple roundtrips, focus the laser pulse to a predetermined position, rotate the plane of polarization by 90 degrees in every completion of the roundtrip; a collection optics which collects lights scattered from the focused region in plasma, ‘first collected scattering’ by the vertical polarization of the laser pulse and ‘second collected scattering’ by the horizontal polarization of the laser pulse; a polychromator which filters the collected lights provided from the collection optics according to spectral characteristics and output the filtered lights; and a computer which measures spectral characteristics of the first and second collected scatterings by using the filtered lights and outputs Thomson scattering signal with the background noise and the background noise without Thomson scattering signal. |
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description | 1. Field of the Disclosure The invention relates to a purifying device for sludge under water and a method for operating the same, and more particularly, to a purifying device operating in a negative pressure, having a small volume with lightweight and being easily carried and backwashed and a method for operating the same. 2. Brief Description of the Related Art A conventional pool for cooling nuclear appliances or storing materials is caused to have radioactive solid particles, such as suspended solids or sludge deposited at a bottom of the pool, spread therein when used to wash, cut and store various radioactive materials. Accordingly, when nuclear appliances are decommissioned or washed or water is purified in a routine schedule, the radioactive solid particles are required to be filtered from the water in order that safety and accessibility of staff members can be ensured and the water in the pool has an enhanced quality. However, a current filtering device has the following drawbacks in use. 1. Currently, the filtering device with a medium or large scale is typically fixed near the pool and is caused to be difficulty in movement without any flexibility to be moved at any time to different polluted regions to be processed thereby. 2. The conventional filtering device is set on the ground, but not under the water. This leads plant areas to be occupied and is not beneficial to protection of staff members during radiation processes. 3. The conventional filtering device operates in a positive pressure such that pumps are subject to being clogged and worn out. The filtering device is required to be equipped with a pressure casing for accommodating a filter. The filtering device after processing the radioactive materials has concerns of safety and radiation protection when maintained and detached. It is possible that the solid particles enter into the inside of the filter due to pressure such that the filter is not easily backwashed. 4. It is complicated to backwash the conventional filtering device, and ways to backwash are not diversified. 5. The filter employed in the conventional filtering device is disposable such that a large amount of secondary waste is produced and is not environmentally friendly. To sum up, the filtering device operating in a positive pressure has the above drawbacks in practice. Accordingly, it is an important issue to lead processes to be simplified with flexibility and working efficiency and to reduce radiation doses to staff members and secondary waste. Thus, the invention is proposed to improve the drawbacks of the conventional filtering device operating in a positive pressure. In accordance with the main objective of the invention, the invention is directed to a purifying device for sludge under water, wherein the device operates in a negative pressure such that pumps or other inner devices can be prevented from worn out and clogged and each parts have enhanced life spans. In accordance with the second objective of the invention, the purifying device for sludge under water is not required to have a pressure casing and is caused to have a reduced weight and volume and to be easily detached under water and maintained. In accordance with the third objective of the invention, the purifying device for sludge under water can expand the number of the filters based on demand and is caused to have enhanced filtering area and filter efficiency. In accordance with the fourth objective of the invention, the purifying device for sludge under water can be moved on a flexible schedule in coordination with an area to be processed and thus has excellent movability. In practice, the device can be mounted in the pool and shielded by the water in the pool such that radiation doses to staff members can be reduced. In accordance with the fifth objective of the invention, the purifying device for sludge under water is convenient due to containing the filter that can be backwashed using filtrates, air and ultrasonics. In order to achieve the above objectives, the invention proposes a purifying device for sludge under water. The device includes a main fixing frame having an accommodating portion assess to the outside, a hollow liquid container arranged in the accommodating portion of the main fixing frame, wherein a liquid-flow hole is arranged at the liquid container, at least a filter arranged on the liquid container, wherein the filter is provided with an outer surface filtering out small particles and a filtrate discharging hole communicating with the liquid-flow hole and a backwash hole at the liquid container, and a pump connected to the liquid-flow hole at the liquid container through a liquid pipeline such that a negative pressure created from the outer surface to the filtrate discharging hole is applied to the filter. In accordance with an embodiment, the backwash hole at the liquid container has a fluid flow therein through a backwash pipe. In accordance with an embodiment, the fluid flowing into the liquid container through the backwash pipe comprises an air, water or filtrate. In accordance with an embodiment, the filter apart from the liquid container is arranged with a plate-shaped pressing part leading the filter to tightly join the liquid container. In accordance with an embodiment, multiple securing parts join the pressing part to the liquid container. In accordance with an embodiment, the main fixing frame has multiple side flanges at a periphery thereof, wherein multiple slings for suspending the main fixing frame are fixed on the side flanges. A method for operating a purifying device for sludge under water includes the following steps: a. moving a liquid container having a filter to an area having a liquid to be filtered so as to fix a filtering device; b. performing a filtering process comprising leading the liquid to flow into the liquid container through the filter filtering out small solid particles contained in the liquid, wherein the small solid particles are attached to a surface of the filter, using a hydraulic pressure of the area and a negative pressure created by a pump extracting a filtrate; and c. performing a filtering determining process comprising determining if small solid particles in the area are filtered out to a tolerance level, wherein performing the filtering process is repeated if small solid particles in the area are determined not to be filtered out to the tolerance level, while all of the processes is stopped if small solid particles in the area are determined to be filtered out to the tolerance level. In accordance with an embodiment, before performing the filtering determining process, the method further includes performing a clogged-filter determining process comprising determining if the filter is clogged, wherein the filtering determining process is performed if the filter is determined not to be clogged, while the method further includes performing a filter backwashing process comprising removing the small solid particles attached to the surface of the filter if the filter is determined to be clogged. In accordance with an embodiment, the filter backwashing process comprises moving the filter to a cleaning tank and leading a fluid in a suitable pressure to flow into the liquid container such that the small solid particles can be removed from the surface of the filter using a positive pressure created by the fluid. In accordance with an embodiment, the filter backwashing process comprises moving the filter to a cleaning tank having an ultrasonic device and cleaning from the small solid particles attached to the surface of the filter using ultrasonic vibration. In accordance with an embodiment, after performing the filter backwashing process, the method includes performing a backwash-liquid collecting process including leading a backwash liquid created in the filter backwashing process to flow into an external deposition collecting container and then performing a collecting process. The accompanying drawings are included to provide a further understanding of the invention, and are incorporated as a part of this specification. The drawings illustrate embodiments of the invention and, together with the description, serve to explain the principles of the invention. While certain embodiments are depicted in the drawings, one skilled in the art will appreciate that the embodiments depicted are illustrative and that variations of those shown, as well as other embodiments described herein, may be envisioned and practiced within the scope of the present disclosure. Illustrative embodiments are now described. Other embodiments may be used in addition or instead. Details that may be apparent or unnecessary may be omitted to save space or for a more effective presentation. Conversely, some embodiments may be practiced without all of the details that are disclosed. Referring to FIGS. 1 and 2, in accordance with a first embodiment of the invention, a device mainly includes a fixing frame 1 having an accommodating portion 11 assess to the outside, wherein the main fixing frame 1 has multiple side flanges 12 at a periphery thereof, wherein multiple slings 121 for suspending the main fixing frame 1 are fixed on the side flanges 12, a hollow liquid container 2 arranged in the accommodating portion 11, wherein a liquid-flow hole 21 and at least a backwash hole 22 are arranged at the liquid container 2, filters 3 provided with outer surfaces filtering out small particles and a filtrate discharging hole 31 communicating with the inside of the liquid container 2, and a pump 4 connected to the liquid-flow hole 21 at the liquid container 2 through a liquid pipeline 41. In accordance with a process flow of the present invention, a method for operating the above device includes the following steps. First, a step P11 of fixing a filtering device is performed by using a lifting tool to move the liquid container 2, on the main fixing frame 1, having the filters 3 to an area having a liquid to be filtered, that is, a pool containing radioactive small solid particles spread therein via the slings 121. Next, a filtering process P12 is performed by leading the liquid to flow into the liquid container 2 through surfaces of the filters 3 for filtering out the small solid particles contained in the liquid, wherein the small solid particles are attached to the surfaces of the filters 3, using a hydraulic pressure of the area and a negative pressure from the outer surface to the filtrate discharging hole 31, wherein the negative pressure is created by the pump 4 driving fluid in the liquid container 2 to be discharged. In the period when filtering process P12 is performed, a clogged-filter determining process P13 is performed comprising determining if the filters 3 are clogged, wherein the filtering determining process P12 is performed if the filters 3 are determined not to be clogged, that is, the filters 3 have a tolerant filtering capability, while a filter backwashing process P14 is performed if the filters 3 are determined to be clogged, that is, the filters 3 do not have a tolerant filtering capability, wherein the filter backwashing process P14 comprises moving clogged filters 3 to a cleaning tank having an ultrasonic device with a backwash pipe 5 joining the backwash holes 22 at the liquid container 2 so as to remove the solid particles from the surfaces of the filters 3 by a positive pressure created by leading a fluid, such as air, water or filtrate, to flow into the liquid container 2, accompanying with ultrasonic vibration created by the ultrasonic device. Thereby, backwashing the filters 3 can be performed. Next, a backwash-liquid collecting process P15 is performed comprising leading a backwash liquid created in the filter backwashing process P14 to flow into an external deposition collecting container and then performing a collecting process. Finally, a filtering determining process P16 comprises determining if solid particles in the area are filtered out to a tolerance level, wherein performing the filtering process P12 is repeated if solid particles in the area are determined not to be filtered out to the tolerance level, while all of the processes are stopped if solid particles in the area are determined to be filtered out to the tolerance level. Referring to FIG. 3, in accordance with a second embodiment of the present invention, the device comprises a hollow liquid container 2a provided with a liquid-flow hole 21a and at least a backwash hole 22a at a peripheral side thereof, and multiple filters 30 arranged at a side of the liquid container 2a, wherein the filters 30 apart from the liquid container 2a are arranged with a plate-shaped pressing part 6 having a square shape same as that of the liquid container 2a. Multiple securing parts 61 join the pressing part 6 to the liquid container 2a, and a pressure created by the pressing part 6 leads the filters 30 to join the liquid container 2a. A method for operating the device is similar to the first embodiment, and the related illustration is omitted. Referring to FIG. 4, in accordance with a third embodiment of the present invention, the device comprises a round hollow liquid container 2b provided with a liquid-flow hole 21b and at least a backwash hole 22b, and multiple filters 30 arranged at a top side of the liquid container 2b, wherein the filters 30 apart from the liquid container 2b are arranged with a plate-shaped pressing part 60 having a round shape same as that of the liquid container 2b. Multiple securing parts 601 join the pressing part 60 to the liquid container 2b, and a pressure created by the pressing part 60 leads the filters 30 to join the liquid container 2b. A method for operating the device is similar to the first embodiment, and the related illustration is omitted. In accordance with the present invention, the device and method have the following advantages: 1. The pump 4 can be fixed under water or at a side of an area or pool to be processed based on different environments and demands. 2. The number of the filters can be easily increased or reduced based on demand such that a filtering area can be altered and filter efficiency can be enhanced. 3. No pressure casing is needed, and the liquid container 2 and the filters 3 can operate under water. 4. The filters 3 and the liquid container 2 can be joined in multiple ways and the arrangement is diversified. 5. Backwashing can be performed using water, filtrate, air or ultrasonics. Accordingly, in accordance with the present invention, the purifying device for sludge under water and the method for operating the same can achieve the following objectives of enhanced life spans of parts, easy detachment under water and maintenance, enhanced filter efficiency due to easy expansion of the number of the filters, and perfect movability. Accordingly, the present invention meets industrial applicability, novelty and inventive steps. Unless otherwise stated, all measurements, values, ratings, positions, magnitudes, sizes, and other specifications that are set forth in this specification, including in the claims that follow, are approximate, not exact. They are intended to have a reasonable range that is consistent with the functions to which they relate and with what is customary in the art to which they pertain. Furthermore, unless stated otherwise, the numerical ranges provided are intended to be inclusive of the stated lower and upper values. Moreover, unless stated otherwise, all material selections and numerical values are representative of preferred embodiments and other ranges and/or materials may be used. The scope of protection is limited solely by the claims, and such scope is intended and should be interpreted to be as broad as is consistent with the ordinary meaning of the language that is used in the claims when interpreted in light of this specification and the prosecution history that follows, and to encompass all structural and functional equivalents thereof. |
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055457983 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to preparing radioactive ion-exchange resins for disposal of their radioactively decaying atoms as waste. Decaying atoms attach to such resins by ion-exchange, for example, as nuclear power facilities clean the water which circulates inside the reactors. This specification teaches methods to reduce the volume of radioactive material which must be stored or buried after use of ion-exchange resins. Exceeding results with present commercial practice in disposal of radioactive ion-exchange resins, this invention provides: (i) removing water and its associated volume from the solid radioactive, ion-exchange resins, PA1 (ii) altering the chemical structure of the radioactive ion-exchange resins to remove ion-attractive groups, thereby avoiding further sorption of water, PA1 (iii) through the removal of the ion-attractive groups, also freeing the original radioactive ion-exchange resins from radioactive ions they had held, thereby forming simple polymer resin, PA1 (iv) depolymerizing simple polymer resin and vaporizing away nonradioactive vapors while retaining radioactive synthetic mineral, PA1 (v) operating in manner in which materials intended to be nonradioactive can be monitored for radioactivity prior to their release, and PA1 (vii) thus allowing safe release of material known to be nonradioactive, thereby reducing the volume of radioactive material that must be stored or buried. PA1 (i) Partial moisture removal and corollary separation of some nonradioactive water from even the solid radioactive ion-exchange resin normally can take place without difficulty. Squeezing, evacuation, and vaporizing are used commercially. PA1 (ii) Mixed hydroxides of sodium and potassium are often good material to add to firmly bind and hold decaying atoms which have attached to the ion-exchange resin. At 1/1 mol ratio and no excess water, these hydroxides fuse at 170.degree. C. If even small amounts water are present, these solutions form liquids at lower temperatures yet retain the ability to firmly bind the decaying atoms. The firmly bound decaying atoms will not escape from the hydroxide environment even if the organic material is chemically separated and removed from the decaying atoms. PA1 (iii) These same hydroxides, particularly if fused, can remove a cation-exchange sulfonyl reactive chemical group or similar group from a benzene ring and form a phenolic group which is neutralized by hydroxide. This replacement is important because it will allow later depolymerization and vaporization of decontaminated fragments of the substrate material of the radioactive ion-exchange resin. PA1 (iv) Heating the radioactive ion-exchange resin will partially depolymerize it. Partial liquefaction will occur both by the depolymerization and by melting of still polymerized segments of linear polymer. Normally the inventor has found it simple and effective to heat gently under air-free conditions which will allow the separational chemical reactions without oxidation. PA1 (v) Along with liquefaction the separational chemical reactions gradually shift to form different fragments as the polymer decomposition moves into the more heavily cross linked regions. As the resin decomposition proceeds, the temperature rises, the color of the decomposition products changes, and the residual solid polymer eventually becomes a charry residue. PA1 (vi) Also, as the ion-exchange resin breaks into the fragments, vaporization of the depolymerized material takes place. This vaporization is important and useful because it separates substantially nonradioactive material from the radioactive residue. PA1 (vii) Pyrolytic degradation breaks bonds in the cross-linked portion of the radioactive resin residue. Most of the degradation products from these separational chemical reactions are volatile at the temperatures used for depolymerization or the often higher temperatures used for pyrolysis. Vaporization is one of the better ways to separate volatile nonradioactive fragments formed here because the radioactive salts are effectively nonvolatile. Often it is useful to operate at less than atmospheric pressure. Other techniques again may be useful in assisting the vaporization, e.g., by steam distillation. PA1 as Taken from the Parent Application PA1 (1) One object of this invention is a method of preparing ion-exchange resin holding radioactive material including decaying atoms for its disposal comprising the steps below. PA1 (1a) At least part of the radioactive material is chemically attached to a bonding material such that decaying atoms become at least in part firmly bonded, whereby parent application first-treated resin residue is created. PA1 (1b) A chemical separation of at least part of the firmly bonded radioactivity from parent-application first-treated resin residue is effected, whereby parent-application second-treated resin residue at least partially freed of chemically attached decaying atoms is created. PA1 (1c) Depolymerizing, at least in part, the parent-application second-treated resin residue, whereby at least partially depolymerized parent-application resin residue is created. PA1 (1d) Bulk physical separation of at least part of the second-treated resin residue from the firmly bonded decaying atoms is effected, whereby substantially nonradioactive parent-application resin residue is created. PA1 (1e) In carrying out the steps above, at least one separation container Is used which will allow retention of at least part of one product resulting from the steps until it can be determined that unwanted release of decaying atoms will not occur as supposedly substantially nonradioactive resin residue is removed for nonradioactive disposal with corollary reduction in the space required for the radioactive disposal. PA1 (2) Another object of this invention is effecting one or more of the steps of the invention at least in part by heating. PA1 (3) Another object of this invention is effecting at least in part one or more steps of the invention in a separation container while the separation container is hermetically sealed. PA1 (4) Another object of this invention is effecting at least in part one or more steps of the invention in a separation container while the separation container is operating at other than atmospheric internal pressure. PA1 (5) Another object of this invention is effecting at least in part one or more steps of the invention at least in part in a separation container while the separation container is operating with an atmosphere in which the thermodynamic activity of oxygen is controlled. PA1 (6) Another object of this invention is pyrolyzing resin residue to break volatile organic fragments from the resin residue under reducing oxygen activity. PA1 (7) Another object of this invention is forming at least some carbon dioxide from substantially nonvolatile carbonaceous residue under oxidizing conditions. PA1 (8) Another object of this invention is using a catalyst in the decomposition of a resin residue. PA1 (9) Another object of this invention is forming and moving of at least one component of a resin residue as a vapor which condenses in substantially nonradioactive form. PA1 (10) Another object of this invention is using at least one type of material comprising metallic oxide to at least in part form said firmly bonded decaying atoms. PA1 (12) Another object of this invention is trapping potential air pollutants on substantially stable and nonvolatile salt. PA1 (13) Another object of this invention is specifically the binding into salt of chemical groups which would complicate later disposal of substantially nonradioactive resin residue by incineration. PA1 (14) Another object of this invention is chemically altering ion-exchange resin holding radioactive material to render it substantially incapable of holding moisture. PA1 (15) Another object of this invention is the use of solvent extraction to separate nonradioactive material from radioactive material by chemical alteration of the original ion-exchange material holding decaying atoms. PA1 (16) A further object of this invention is to monitor a separated phase while it is still in containment in order to assure it is substantially nonradioactive. PA1 (17) A further object of this invention is to use a technique to assist transport of organic vapor to a condensation region of a separation container. This invention is urgently needed: First, most commercial nuclear power plants in the United States have already lost all access to any burial for their radioactive wastes--such wastes must be stored. Also, most other commercial operations which generate radioactive waste are faced with an uncertain period of storage as their wastes accumulate. Without storage space most of the commercial operations indicated would have to close down. (Later note: The Barnwell burial site reopened Jul. 1, 1995.) Long-term radioactive storage of radioactive wastes was being planned, for example, at the Perry nuclear power facility near Cleveland in October, 1992. Both State and Federal new burial facilities were supposed to be prepared: Federal law once mandated that states would have to supply radioactive burial sites, but the requirement was overturned by the U.S. Supreme Court; litigation continues. The Federal burial site for commercial radioactive waste was supposed to be available in 1998, but estimates say it is 15 years behind schedule. Second, open Federal sites for burial of radioactive wastes are rapidly filling while waste generation continues, and there are strong objections by U.S. citizens to any burial or transportation of radioactive materials. Third, environmental logic requires that radioactive burial volumes be minimized. Lacking the teaching of this invention, current Federal practice is to bury considerably more waste than would be buried with improved practice as described in this invention. For those organizations which must store their radioactive wastes, excessive storage is illogical both environmentally and economically. 2. Description of the Related Art As noted above, decaying atoms in water are often removed onto ion-exchange resin. In much industrial practice, and presumably also widely at Federal facilities, the radioactive ion-exchange resin is packaged wet in drums for storage or disposal. Because steel drums rust, concrete reinforcement was added for some physical protection against radioactive leakage. Current practice often uses glass-reinforced plastic drums with no interior reinforcement against their damage. Other than to remove some of the water, the resin characteristics are not changed before storage or burial. Such resin, if exposed to weathering, can release radioactive atoms it holds. Long-Term Burial: As noted above, long-term burial as used in most past practice is not now an option for most commercial generators of radioactive waste. Federal burial grounds are filling up, and Federal generators of nuclear waste are facing many future problems with burial, particularly excessive burial. Waste-volume reduction is needed. Burial has always been considered a problem. In the inventor's experience from 1946 and still continuing, there has been concern that much buried radioactive material would have to be dug up and moved. Times and environmental concerns, as well as standards for acceptable burial, have changed, both as to form and volume of materials which are acceptable. Ion-exchange resins have long been considered a special problem because they can pick up and hold large volumes of water of hydration, swelling in the process. Open-Flow Incineration: The term open-flow incineration is used here for typical incineration such as is used in incinerating either garbage or wastes of paper and plastic. Here oxygen, usually in air mixed with other gases, flows over hot material and reduces the material substantially to ash. Typically, water vapor and carbon dioxide are the principal gases formed. Other gases, e.g., noxious oxides of nitrogen and of sulfur, may form. Bits of the ash dust typically will be carried along with the flowing gas. Traps to remove the gaseous oxides, plus filters to remove the dust, can be installed along the flowing-gas path to the stack. Most of the time these traps work well, e.g., when such systems are used to burn mildly radioactive paper and rubber gloves, which generate ash. Open-flow incineration systems neither (i) hold the gas for precise analysis for carried radioactivity before the gas is released to the atmosphere nor (ii) stop the incineration instantaneously if excessive radioactivity is detected in material escaping up the stack. One learns too late that something has gone wrong and uncontrolled decaying atoms are escaping. A large incinerator at is planned Oak Ridge, Tenn., for commercial nuclear waste. Discussions by the inventor with incinerator personnel suggest that the facility will not be suitable for ion-exchange resin for reasons discussed below. Incineration of Radioactive Ion-Exchange Resin: In addition to the incineration problems noted, radioactive ion-exchange resin lacks ash-forming materials to trap the radioactive dust released as incineration occurs. This dust, if not trapped, may be expected to be blown around by the gas stream. Also, a significant fraction of the resin volume is as inorganic chemical groups which were put there to trap ions. Incineration releases chemically nonradioactive but noxious gases which must be trapped for environmental reasons. Trapping the noxious gases and the radioactive dust by conventional technology, even if the technology were to work perfectly, might actually increase the volume of radioactive waste to be stored or buried. For these and other reasons, burial is widely preferred over open-flow incineration for disposal of radioactive ion-exchange resins--incineration often is not a good choice. Because a dictionary definition of incineration involves "reducing something to ash," it is noted that incineration, as used in this disclosure, includes oxidation of carbonaceous residues in the vicinity of radioactive oxides or other salts to remove the carbon as carbon dioxide. The treatment of this invention is not an open-flow system--rather, all gases are trapped and held available for radioactive monitoring before they are released. Pyrolysis of Radioactive Ion-Exchange Resin: It is noted that pyrolysis is often combined inherently with incineration because of normal lack of local oxygen at heated combustion regions. Such normal pyrolysis fails to utilize the concept of depolymerization, followed by pyrolysis, if that is required, as offered by the present invention. With more control of the chemical bond breakage, one can (i) depolymerize ion-exchange resin, (ii) meanwhile break off large organic fragments from the depolymerizing resin, (iii) thereby vaporizing mostly condensible vapors, and (iv) condense these vapors and monitor the condensate for radioactivity. Over 95% reductions in the volumes of potentially radioactive gases generated may be achieved with the present invention, as compared with use of normal incineration practices. Aqueous Oxidation: Processes are being developed that employ hydrogen peroxide to oxidize ion-exchange resin to carbon dioxide, water, and derivatives of sulfonyl and trimethyl amine groups. As compared with the present invention, aqueous oxidation, like open flow incineration, generates very large volumes of potentially radioactive gas. With aqueous oxidation, the gas is generated in radioactive water which may become entrained in continuous gas flow. Such flow may lead to very finely divided, highly radioactive particles that, when dry, can be carried in even gentle winds. Also, the system must be treated to handle sulfates and radioactive materials after the ion-exchange resin has been destroyed. The peroxide may also convert radioactive cations to anions, which may be harder to collect and dispose of than were the original anions. With the present invention, in contrast, sulfates formed from the cation-exchange resin may become part of synthetic minerals, and anions present may become cations that coprecipitate readily inside the synthetic minerals. Such minerals have much better anticipated lives for protecting against release of decaying atoms than do steel, concrete, or plastic, as now used. Other Methods of Decontamination from Decaying Atoms: Numerous other decontamination methods might remove and isolate decaying atoms from a source, e.g., coprecipitation alone, solvent extraction, vaporization, and leaching. For solid radioactive material such as an ion-exchange resin, however, most of these techniques are substantially inoperable because the nonfluidity of the solid effectively blocks thorough removal of the decaying atoms in the interiors of solids. Many customary techniques for handling solids such as metals or oxides use aqueous solutions to dissolve them. Such solutions can then be subjected to near-equilibrium separations processes. However, unless there is resin destruction, aqueous dissolutions are largely inoperable for solid radioactive ion-exchange resins. Summary Regarding Related Art: The existing art for storage or burial of radioactive ion-exchange resins involves excessive volumes which are environmentally and economically unsatisfactory. Likewise, the concepts of existing art for resin destruction appear to be environmentally and economically less satisfactory than are the concepts of the present invention. Patents Noted: Buchwalder, et al., U.S. Pat. No. 4,122,048, used a basic compound to block the active sites of certain contaminated ion-exchange resins so that these resins could be encapsulated in further resin for disposal. The procedure neither offers long-term environmental protection nor reduces the radioactive volume to be disposed of. Laske, et al., U.S. Pat. No. 4,732,705, added various chemicals to reduce the swelling upon wetting of ion-exchange resins. This treatment may reduce the disposal volume of the resins, but it does not offer long-term environmental protection and may actually tend to release the radioactive ions the resin initially held. Knotic, et al., U.S. Pat. No. 4,235,738, added high-boiling oil to ion-exchange resin prior to its heating to produce decomposition of the resin by carbonization. This treatment may assist in retaining the decaying atoms, especially by lowering the carbonization temperature, and avoiding some vaporization of decaying atoms. However, the carbonaceous material formed (i) fails to offer long-term environmental protection of the entrapped decaying atoms, and (ii) the carbon present during carbonization tends to increase the decomposition and vaporization of materials such as radioactive cesium oxide. Kawamura, et al., U.S. Pat. Nos. 4,636,335 and 4,654,172, use low temperature pyrolysis to separate ion-exchange groups from ion-exchange resins prior to high temperature pyrolysis. Then the hot resin residues are compressed into a "molded article". They note, "In this way, decomposition gases generated during thermal decomposition are separated in two stages and gaseous nitrogen oxides (NO.sub.x) and gaseous sulfur oxides (SO.sub.x) which require careful exhaust gas disposal treatment are generated only in the first stage thermal decomposition . . . " ('335, column 2). This Kawamura, et al., preliminary procedure reduces the volume of gas initially produced and yields a carbonaceous residue that provides largely physical, rather than chemical, trapping of the decaying atoms. However, the '172 claims 7-9 also note "presence of a vitrifying agent which absorbs volatile radioactive substances" that were "added before the pyrolysis at a low temperature" such as glass frit. A frit has substantially no contact with most of the decaying atoms, and it therefore cannot pick them up. The '335 and '172 treatments (i) do not chemically anchor the decaying atoms in a condensed phase, i.e., as solid or liquid, prior to vaporizing resin components, (ii) do not afford dependable environmental protection against release of many radioactive elements if the hydrocarbons of the carbonaceous residue have become oxidized by air or otherwise, and (iii) do prevent precise reversal of the polymerization reactions which originally formed the ion-exchange resin. SUMMARY OF THE INVENTION This invention offers a new method for assisting in preparing ion-exchange resin holding decaying atoms, i.e., radioactive ion-exchange resin, for its disposal by reducing the volume of radioactive material which must be stored or buried after use of the ion-exchange resin to remove decaying atoms from radioactive water. Before describing the concepts of the invention, it is useful to discuss the nature of ion-exchange resins in general and radioactive ion-exchange resins which are of particular interest here. The Starting Nonradioactive Ion-Exchange Resin, Its Manufacture, and Some of Its Reactions: First, recognize that an ion-exchange resin is designed for either capture of cations or of anions, i.e, respectively, like Na.sup.+ on cation-exchange resin or Cl.sup.- on anion exchange resin. In this invention the chemical treatments are primarily directed toward the cation-exchange resins, but the procedures to a large extent also lead to capture of the anions which were initially present, as is further discussed later. A typical starting material for making ion-exchange resin will be what is often called polystyrene. It is in a class of polymers that are called synthetic resins. Before polymerization, the styrene (C.sub.6 H.sub.5 --CH.dbd.CH.sub.2) usually will have been mixed with about 8% of divinyl benzene (CH.sub.2 .dbd.CH--C.sub.6 H.sub.4 --CH.dbd.CH.sub.2), which causes cross-linking of the styrene/divinyl benzene chains during polymerization. During polymerization, the double bonds shown above break to forms chains of mixed styrene and divinyl benzene, as indicated for styrene chains in Equation 1: ##STR1## This polymer is not yet an ion-exchange resin--reactive chemical groups must be added with different groups being effective for attachment of cations or of anions. The polymer resin, often as beads or grains, must have been treated further. Either cation-exchange groups, e.g., sulfonic acid groups, which hold cations, or anion-exchange groups, e.g., quaternary ammonium groups, which hold anions, are added. The sulfonic acid group attaches to carbon on a benzene-type ring of a polymerized styrene or divinyl benzene, while water is given up to concentrated sulfuric acid (HOSO.sub.2 OH) as represented below; ##STR2## represents a styrene in a polymer chain: ##STR3## This is the hydrogen-ion form of the polystyrene cation-exchange resin. It readily gives up the hydrogen ion in exchange for other inorganic cations. The sodium ion exchange forms sodium sulfonate: ##STR4## For Ba.sup.++, two sulfonyl sites are converted to barium sulfonate forms: ##STR5## Usually the higher charged cations are held more strongly. These bonds involving the sulfur are not yet referred to as "firmly bonded" because of the relative weakness of the C--SO.sub.3 bond as compared with completely inorganic bonds, e.g., in BaSO.sub.4. Bonds are discussed further below. Radioactive Ion-Exchange from Nuclear Power Reactors: In the case of pressurized water nuclear reactors or boiling water nuclear reactors, most of the radioactive ions of decaying atoms are cations from corrosion of the metals in alloy containers for the water flow, but anionic species can also be present. Radioactive ions of cobalt, zinc, manganese, chromium, cesium, iron, technicium, antimony, iodine, hydrogen, carbon, and other elements may be present. Waste resin drums from nuclear power stations may give off 0.8 to 80 R/hr of nuclear radiation as registered on a hand monitor. These radioactive ions attach to the ion-exchange resin to form radioactive ion-exchange resin, which is the material whose radioactive volume this patent seeks to reduce. The attachments by the radioactive ions are analogous to those by Na.sup.+ and Ba.sup.++, and the equations describing the cation-exchange resin behavior are like those for Na.sup.+ and Ba.sup.++, Eqs. 3 and 4. Both anions and cations of the metals appear to be amenable to treatment by the present invention. Concepts of Use in the Invention: Thermodynamic data show that organic hydrocarbon compounds such as polystyrene resin are generally weakly bonded in a chemical sense, as compared with the firmly bonded structures of many inorganic substances. For example, weakly bonded carbon-to-carbon attachments n polystyrene resin may break spontaneously in an inert atmosphere at 300.degree. C. Such broken attachments may reform or form new linkages. Corollary resin decomposition will sometimes form gases, e.g., methane, and vapors, e.g., styrene and even larger molecules such as styrene dimer. The proportions of different compounds in vapor mixtures are influenced by numerous factors, e.g., heating rates and temperatures. In contrast with the hydrocarbon compounds, many inorganic crystals are firmly bonded, e.g., barium sulfate, which can be heated at 800.degree. C. in an inert atmosphere without significant breakage of its bonds. Likewise, anhydrous sodium sulfate is firmly bonded and can be heated to high temperatures. Furthermore, sodium sulfate dissolved as hydrated ions in water is also firmly bonded--the sodium sulfate would not have dissolved in water if it had not become even more firmly bonded in solution than it was as the anhydrous form. The solutions can be dried back down to anhydrous sodium sulfate. Resin decompositions at temperatures in the range 150.degree.-500.degree. C. are affected by the presence of at least some other materials. For example, anchor materials that are selected primarily to assure that radioactive atoms will become permanently trapped for permanent disposal may also lead to formation of resin-decomposition catalysts. As in experimental Cases 1 and 2, discussed later, it appears that such catalysts can focus the breaking of carbon-to-carbon attachments to achieve resin decomposition by depolymerization, giving primarily styrene and divinyl benzene. Simple pyrolysis gives a more complex spread of products. Directed energy matching a particular bond strength may also be useful, e.g., using electromagnetic radiation that can add energy to, and break open, a particular type of bond. As examples, one might irradiate the radioactive ion-exchange resin with an energy which would readily break a type of bond at which one wishes to have reaction occur, e.g., to free substantially all radioactive material and sulfonic groups from an organic residue. Catalysis suitable for efficient depolymerization of the organic polymer resin that has been freed from its radioactive material appears to occur with barium compounds. The presence of barium hydroxide, barium sulfate, or both, as the resin-decomposition catalyst experimentally led to large fractions of depolymerization with low fractions of relatively noncondensible gases and charry residues. This situation is valuable in operation of this invention. Critical actions of anchor materials are to supply ions that bond to and anchor ion-exchange groups such as sulfonyl groups and to assure that most types of decaying atoms present will remain with the anchored ion-exchange groups. Eventually these decaying atoms and anchored sulfonyl groups will become firmly bonded radioactive material, e.g., radioactive synthetic barite. One can first attach sulfonyl groups of a cation-exchange resin to anchoring ions from anchor material, e.g., Ba.sup.++ from barium hydroxide, thereby forming barium sulfonates. With the sulfonate groups' bonds so anchored, it becomes possible to create conditions favoring chemical reactions that separate these groups from polymerized organic matter to which they had been attached. In these reactions the sulfonate groups in most cases become part of an inorganic sulfate; in some cases sulfite might also form. Meanwhile, the organic portion of the original ion-exchange resin becomes chemically free of, though mixed with, the radioactive material. The amount of condensed-phase residues from resin decomposition, such as tarry materials and carbonaceous solids, appeared to increase with the release of gases or vapors other than styrene or divinyl benzene. The interactions among carbon atoms in condensed-phase residues may produce firmly bonded structures in the sense that the residues do not undergo much thermal decomposition even at higher temperatures. Chemical interactions of such resins with inorganic materials are, in most cases, very weak. These condensed-phase residues are not capable of firmly bonding to inorganic species such as cations or compounds of decaying atoms. However, these elements, which had earlier attached to the sulfonic acid cation-exchange resin, might become physically trapped for some time, e.g., until the tars oxidize away during burial or storage and allow the decaying atoms to escape. Attachments of polystyrene to sulfonyl or quaternary ammonia groups are particularly weakly bonded. Some release of these groups can be achieved by heating ion-exchange resins at less than 300.degree. C. for example. The novel group of steps which comprise this invention are based in part on understanding of the chemical concepts above. Unobviousness is evident from existence of the problem of excess burial volumes in disposal of radioactive ion-exchange resins that has existed for over forty years. The Broad Concept: The letters in parentheses in the following discussion correspond with those in Claim 1. The central concept of this invention is to allow reaction among (a) radioactive ion-exchange resin that includes decaying atoms and cation-exchange resin, (b) anchor material that can supply anchoring ions that can react at least in part with the decaying atoms and the cation-exchange resin, and (c) water in some form. These materials (d) are brought together where they can react. Usually the initial reactions are at room temperature. Included among various possible activities of the water are forming hydrated ions, acting as a medium in which reactions may take place, and resupplying reactant H.sub.2 O which was generated and removed during manufacture of the cation-exchange resin. This H.sub.2 O resupply may be useful prior to decomposition of the ion-exchange resin, as discussed below. One reaction is (e) the attachment of anchoring ions to the cation-exchange resin. These anchoring ions are supplied by the anchor material, typically through the water, to the cation-exchange group on the resin. This attachment replaces the hydrogen ions on the resin with anchoring ions, but the cation-exchange group remains attached to the resin, e.g., typically a sulfonate group on polystyrene, as discussed earlier. Anchored cations on first-treated resin are formed. Also, (f) the anchoring ions provide an aqueous ionic environment in which radioactive ions are held by charge interactions. Whether anions or cations, and whether the species are in aqueous solution or are on cation or anion resin, these ions cannot readily escape even if the resin is being destroyed or, later, being removed. Anchored decaying atoms are created. Next, (g) bonds from a cation-exchange site to an organic portion of the resin are exposed to reaction by supplying energy and a third portion of anchoring ions at points where organic/inorganic bonds join organic portions of the first-treated resin to the anchored cation-exchange groups. Because the anchoring ions have attached with strong bonds to, for example, form a sulfonate group, the attachment of the carbon of the resin, i.e., of the organic polymer, to the sulfonate group has become more vulnerable to attack, and such an attack may become highly selective. Once an organic/inorganic bond has been prepared for reaction, it becomes possible for (h) the anchored cation-exchange groups to attach additional anchoring ions and convert, for example, a sulfonate group to inorganic sulfates or sulfites. If cation-exchange groups other than sulfonate groups are present, they also in most cases will be converted to similar inorganic compounds. Such inorganic materials are firmly bonded, both as the major components and as the radioactive ions the major components hold. These inorganic materials are at least in part chemically freed from organic material. If water reacts at an organic/inorganic bond at the time other reactions are taking place, this will allow reversal of the sulfonation reaction that was carried out during manufacture of the cation-exchange resin. This sulfonation reaction involved water removal to concentrated sulfuric acid and formation of the sulfonyl groups. With regeneration of the sulfate group by the water reaction, it is possible to form principally sulfates, e.g., BaSO.sub.4. These sulfates, and sulfites, if present, are readily separable from the organic material even though they are physically mixed with organic material. Once the inorganic material has formed, (i) the organic polymer residue is also chemically freed from the anchored cation-exchange groups. Depending on what has happened at the organic portion of the organic/inorganic bond, a number of reactions may take place. With the water addition mentioned, polystyrene may have reformed. Without the water addition, there is a hydrogen shortage in the organic region, and other species presumably will have formed. With organic and inorganic materials physically mixed, (j) any of a number of physical separations would potentially be useful: The preferred embodiment assumes approximate conformance to a two-step separation in which the "polystyrene" resin first depolymerizes to styrene and divinyl benzene, then these materials vaporize away to condense as materials which are either already nonradioactive or can be made so. Even without vaporization, if sufficiently heated the resin can liquefy by a combination of factors such as direct melting and dissolution of the polymer in styrene and divinyl benzene or their small aggregates such as dimers, etc. Also, other solvents could be added to assist the polymer dissolution. Once the organic polymer residue became largely liquefied, it could be filtered or decanted away from an inorganic residue such as BaSO.sub.4 residue rather than requiring vaporization as in the preferred embodiment. Overlapping of the Steps: It is not assumed that these steps will be individually observable. For example, on a microscopic scale the method may be conceived of as successive steps of separating substantially nonradioactive material from a radioactive ion-exchange resin while retaining the decaying atoms in smaller and smaller volume. However, the steps may be largely conceptual. For example, an intermediate step of melting may, or may not, be identifiable when depolymerization, vaporization, and sublimation of organic vapors take place at solid/liquid mixtures of hot, partially depolymerized resin. However, the existence of some sort of melting is important in opening the ion-exchange resin to reaction. It is important to recognize that, on the bulk scale in commercial operations, these steps routinely will take place at different times in different portions of the resin. All the steps listed are believed to be consistent with the inventor's experiments and other somewhat related experiments of which he is aware. Variations within the Broad Concept: Formation of firmly bonded radioactive material including other elements from the group consisting of Groups IA, IIA, and IIIB of the periodic table are noted as sources other than barium hydroxide and NaOH-KOH mixtures. Other anchor materials might be used to provide hydroxide. Air is normally excluded in steps g to j in the section on The Broad Concept above to prevent cation oxidation to anions. Inert gases may be used to displace the air. Energy must be supplied as described in step g in the section on The Broad Concept above. Both heat and electromagnetic energy may be useful, alone or together. Application of this energy may allow water to react chemically at the opened bonds. Such reaction may effectively reverse the sulfonation reaction used during the manufacturing of the starting sulfonated resin. Firmly bonded synthetic barite, BaSO.sub.4, forms as the radioactive ion-exchange resin is separated chemically into organic and inorganic fractions in the preferred embodiment. The barite formation also causes precipitation of radioactive ions and encases these decaying atoms that had been held on the radioactive ion-exchange resin. The decaying atoms, as they are released from organic attachment, may simply attach to the barite and be engulfed, but usually there is also coprecipitation in which Ba.sup.++ and SO.sub.4.sup.= sites are occupied by radioactive ions. For examples, one may choose to think of FeSO.sub.4 from Fe.sup.++ and BaCrO.sub.4 from CrO.sub.4.sup.= in solid solution in the BaSO.sub.4 host. Thus both anions and cations of the radioactive elements of most interest at boiling water reactors can be accomodated in the barite. Furthermore, the reduction of many anions by hot organic matter prior to bulk formation of the barite will lead to most radioactive elements being present as cations. After the formation of bulk barite, air cannot reach the radioactive elements because they are almost totally within the barite crystals' ionic lattices. Both the synthetic barite and the radioactive ions that it holds are considered to be firmly bonded, i.e., the bonds are strong enough so they cannot readily be broken. Decaying atoms in NaOH-KOH mixtures or the corresponding sulfates, along with similar compositions including elements from the group consisting of Groups IA, IIA, and IIIB of the periodic table, are also firmly bonded. Depolymerization of the organic polymer residue can be used at least in part to form depolymerized residue prior to physical separation of organic material from the firmly bonded radioactive material. Relative to solid polymerized resin, the depolymerized residue may be largely or entirely liquid and may have largely components that are readily volatile. The bulk physical separation may be achieved at least in part by vaporization with corollary transport to condensation elsewhere of the depolymerized residue. The effect is to create vaporization residue, if vaporization is not complete, plus vapor transported organic material. Vaporization and vapor transport may be assisted by the flow of an inert carrier gas that carries components of depolymerized resin as vapor at less than atmospheric pressure; such flow allows major vapor movement at less than the atmospheric boiling temperature. Portions of a vaporization residue may be further removed by pyrolysis or oxidation, either or both. As noted earlier, radioactive anions that have been heated above room temperature may be reduced to cations by reaction with organic materials. Such reaction can occur at lower temperatures but is normally strong at temperatures where chemical separation of firmly bonded radioactive material from organic polymer residue takes place. Bulk physical separation of firmly bonded material and liquefied organic polymer residue may also be achieved by filtration or decantation that pass the liquid and retain the firmly bonded material. Although highly efficient separations are normally most useful, even retention of only 75% of the radioactive material present may be useful for some types of decaying atoms. The present invention was designed to allow retention of all separated materials until they had been monitored for radioactivity. This approach avoids a common problem met by incinerators and other units that release large volumes of radioactive gases flowing continuously. Such units have periodic releases of radioactive material to the atmosphere when the filtration system breaks down. In contrast, the present invention provides that (i) any problems in the retained organic materials can be detected and corrected before there is release, (ii) gas volumes are very small because large organic molecules are vaporized, and (iii) very few noncondensible gases are formed. If unwanted radioactivity is detected, the material can be cleaned up before it is released. As with organic/aqueous solvent extraction, an aqueous wash, e.g., with dilute acid, can remove most possible radioactive contaminants from organic materials which have been retained for radioactive monitoring. If decaying atoms are detected, most will have been physically carried in the moving vapor, and the aqueous environment will be more favorable to them than will the organic. Usual anion-exchange resin would release trimethylamine during the course of this invention. This material could collect in the vapor transported organic material. Acid washing would remove the trimethylamine as a dissolved salt. Treatment of Radioactive Ion-Exchange Resins in the Parent Application: In the parent application for this continuation-in-part, mixtures of NaOH and KOH were the preferred chemicals for making possible this invention's separation of radioactive ion-exchange resins into radioactive and nonradioactive portions--physical separations are made of radioactive material holding decaying atoms and other material which could be disposed of on a nonradioactive basis. However, Ba(OH).sub.2 .cndot.8H.sub.2 O now provides the preferred embodiment for the separation of this invention and has been emphasized. The following discussion of the NaOH-KOH mixtures has been retained with small modifications to save the historical record of the parent application. Reduction of the Radioactive Volume As Described in the Parent Application: To achieve the volume reduction for radioactivity from radioactive ion-exchange resins, one typically goes through several processes. The processes listed separately below are often going on simultaneously. They lead to effecting various steps of the claims made. Other processes may also be used and not all processes are necessary: Complete water removal requires resin alteration. Partial water removal must be considered temporary unless further action is taken to destroy the ability of the radioactive ion-exchange resin to again sorb water. On drying of sodium and potassium hydroxide which have picked up sulfate (see next paragraph) and hold decaying atoms, the decaying atoms will be held as oxides or other salts mixed in the otherwise nonradioactive bulk. They will not be dusty. If desired, the hydroxides can be neutralized for long-term storage. The hydroxide can also release, for example, trimethyl amine from a quaternary amine anion-attracting reactive chemical group and leave a --CH.sub.2 OH group on the benzene ring. The trimethyl amine or its decomposition products can then escape as gas and be trapped in water or acid. Thus, the hydroxide addition can prepare the system for depolymerization, vaporization, and controlled pyrolysis as will be discussed. Depolymerization leads apparently to some, but not complete, unbonding of the polystyrene and other chains. Regarding the depolymerization, recognize that the polymer initially produced was changed to form the ion-exchange resin. Therefore, the depolymerized materials will be modified relative to the original materials which were polymerized. Vaporization aids are useful in retaining large, nonradioactive, organic fragments. Here water vaporization can provide elements of steam distillation. And lowered pressure can let the fragments boil at lower temperatures. For a cross-linked ion-exchange resin like those made from styrene-8% divinyl benzene, slowly raising the temperature can break more and more bonds and release more and more volatile fragments until finally a charry residue is left. Recognize that the charry residue will also hold remains of reactive chemical groups such as sulfonic acid and perhaps quaternary amines on oxides or other salts. From the radioactive ion exchange resins, decaying atoms will be imbedded in the charry residue. These decaying atoms are not firmly bonded, however. Objects of the Invention with Explanations Various steps in the method may in some cases take place substantially simultaneously. While the steps are described with use of well known terms for different types of chemical reactions, to optimize the effects of these reactions they should be carried in specialized ways as taught in this section, in the description of the preferred embodiments, and elsewhere in the specification. "Bonding material", as used with this section of the parent application, is replaced elsewhere in this continuation-in-part by "anchor material" and "anchoring ions", which are derived from anchor material. "Firmly bonded" requires that the decaying atoms will remain substantially in a nonvolatile form in a condensed phase (liquid or solid) with the bonding material even when organic materials to which it has been attached (through an inorganic group) are breaking free of the resin, of the radioactivity, or of both. Firmly bonded is restricted to inorganic bonds. The bonds of ion-exchange resin to the decaying atoms are not broken all at once, so the reactions to attach the decaying atoms to the bonding material should be carried out gently. Too vigorous reaction may prematurely break bonds, spatter liquid solutions and carry decaying atoms in several ways, e.g., in droplets, as solids, in decaying atoms still attached to organic fragments, etc. Carried decaying atoms may contaminate the system where it should be free of radioactivity. With the precautions taught in this specification, and with experimental preparation to learn the behavior of the particular ion-exchange resin system involved, the inventor's experiments have shown that firmly bonded decaying atoms can be formed without substantial transport of decaying atoms. Many metallic oxides form suitable firmly bonded decaying atoms. The inventor has found that mixed sodium and potassium hydroxide have special usefulness in several ways: Molten hydroxides or hydroxide solutions can be used as mobile and readily reactive liquids. The liquids can be contacted with radioactive organic phases to attach both to anionic and cationic decaying atoms. They can also attach to inorganic groups which are chemically attached to resins to create ion-exchange resins. Glass powder may also be a useful oxide which can be made fluid. Other oxides, usually as powders, and other reactive chemicals, can be used similarly to attach to decaying atoms or inorganic resin groups. Other molten salts and aqueous solutions are examples of other sources to firmly bond radioactivity. Heating to effect the chemical separation is a preferred method. Other sources of energy are also potentially useful, e.g., radiation, ultrasonics, or oxidation-reduction reactions. With ion-exchange resin one must be careful in this chemical separation step. One should be confident the firmly bonded decaying atoms either have formed or will be formed as the parent-application first-treated resin and parent-application second-treated resins are also formed. Specifics of this treatment for various possible ion-exchange resins and forms of decaying atoms should be studied experimentally for best performance of a separation unit. For this chemical separation step, poorly miscible radioactive and nonradioactive components may remain physically mixed or even dissolved, but the decaying atoms should not remain chemically on the resin residue. In particular, in the event of separation of radioactive and nonradioactive phases, the decaying atoms will substantially follow bonding material rather than the resin residue. The chemical separation often may also usefully remove ion-attracting chemical species from the ion-exchange resin, thereby destroying the ability of the resin to hold radioactive ions. Again the precautions just mentioned regarding gentle treatment and experimental studies of the particular system will hold. Removal from the ion-exchange resin of sulfate precursors and of nitrogen species along with decaying atoms by the bonding material is particularly notable from an environmental standpoint. These three pollutants create key problems with incineration of radioactive ion-exchange resins and have worked to make incineration of ion-exchange resins largely impractical. In addition, the major driving force for water sorption and retention by the ion-exchange resin is the establishment of an osmosis-like equilibrium involving sorbed ions on the resin. Removal of the ion-exchange component of the resin greatly reduces the resin's capacity to hold water. Here different radioactive ion-exchange resins with different attached and sorbed ions will behave differently toward moisture, and the appropriate chemistry should be evaluated theoretically and experimentally. Depolymerization is dependent on conditions in the system. The inventor has found that partial evacuation while heating the ion-exchange resin or resin residues is useful if used in moderation. If moisture is present, evacuation of the heated mixture will largely remove the moisture. Also, it will assist vaporization of large nonradioactive organic fragments from the resin residues. Too much evacuation can lead to excessive volumes of gas flow plus boiling and bumping. Corollary physical transport of decaying atoms in liquid droplets may occur. Again the teaching of this invention should be heeded, and experimental studies should be carried out prior to operating commercially. Polymerized resin is solid, though porous, and has chemical similarities to synthetic rubber. As such it will resist treatments to separate its decaying atoms from the bulk material, and its resistive character must be destroyed. The inventor prefers depolymerization to the extent possible to turn the hot solid largely into a liquid. Polymerized resin is also capable of holding large amounts of water if the conditions are suitable. Problems with this water retention are discussed elsewhere. As the process of this invention has developed following the inventor's experiments, depolymerization has allowed removal of large fractions of the original ion-exchange resin. The fractions removed normally include separate phases of water and of nonradioactive organic materials, most of which can be largely separated away from nonvolatile radioactive residues. The condensed vapors from depolymerization are potentially disposable as useful chemical feedstocks or as nonradioactive wastes which can be incinerated by usual techniques. Depolymerization of the second-treated resin residue also may create largely immiscible liquid solutions suitable for aqueous-organic solvent extraction if that technique is to be used for radioactive separations. Heating rates of the resins and residues influence the amount of char formed in the resin residues, and the specific resin behavior should be studied theoretically and experimentally. The inventor's experiments with NaOH-KOH bonding material also show that the cross-linkage portion of the resin (often about 8% cross-linked) will not necessarily depolymerize, but this portion can be pyrolyzed to give further decomposition of the original resin. In the inventor's experience in working on this invention, it is preferable to use vaporization and condensation to effect the physical separation. In commercial practice, once an engineer understands the techniques here taught, and assuming use of a suitable separation container built to conform to these teachings, the separation is technically possible and will not be unduly difficult to effect. With the preferred embodiment as tested at bench scale by the inventor, the vaporization and condensation have given excellent separation of nonradioactive moisture and organic fragments from a radioactive residue. Other techniques of separation could be used, e.g., aqueous-organic solvent extraction. Again here the conditions under which the chemical steps have been taken may infuence the nature of the materials being solvent extracted. Separation containers used for the preceding steps should be capable of substantially being sealed, evacuated, pressurized, heated, loaded, and unloaded. They should be sufficiently resistive to reaction with the container contents. They should allow separation of various chemical fractions such as chemical reactants from various products. They should allow measuring, sampling, analyzing, and chemically treating of the container contents in locations where they are collected. Most often in the inventor's experiments resistance heaters, natural gas combustion, or electronic ovens have been used as the heat sources. The control and retention of decaying atoms until nonradioactive portions of separated materials can be monitored is a critical aspect of this invention. Hermetic sealing is one preferred method of such control. As noted above lowering the pressure often beneficially increases the fraction of large, nonradioactive gaseous molecules evolved during depolymerization or pyrolysis of the resin residue. Raising the pressure in the container may beneficially assist the condensation of gases which have been liberated and are to be condensed. Control of oxygen activity is important, for example, in the decomposition of the resin. Under reducing conditions the pyrolysis leads to vaporization of relatively large, substantially nonradioactive organic species which can subsequently be condensed in cooler portions of the vessel. With oxidizing conditions following the pyrolysis, carbon dioxide and moisture can form. The moisture is usually readily condensible; the carbon dioxide may require both pressure and cooling to get it to condense for monitoring before releasing it in nonradioactive form. Oxygen activity also is important in other ways. The inventor's experiments show that pyrolysis of resin residues can be made to form largely condensible, nonradioactive vapors. Residual carbonaceous residue which forms can be crushed readily and does not hold significant amounts of water. The carbonaceous residue which may remain along with the firmly bonded decaying atoms after pyrolysis may trap some decaying atoms which may be disposed of as radioactive material, if no other treatment is used. Heating rates, pressures, and temperatures alter the character of the carbonaceous residue. Formation of carbon dioxide may be disadvantageous in the early steps of the claimed invention, as discussed regarding the fifth object of this invention. Specifically, carbon dioxide formation may (i) excessively raise internal pressures in a separation container, (ii) hinder vapor transport of larger organic molecules to condensation sites after these larger molecules have been separated from the firmly bonded decaying atoms, and (iii) create gas volumes which are difficult to hold until they have been monitored to assure they are substantially free of decaying atoms. The inventor's experiments, however, show that formation of carbon dioxide in later stages of the invention can be useful in removing residual carbonaceous chars from radioactive residues of resin decompositions. Catalysts such as oxides of copper and manganese can assist in the formation of carbon dioxide, and a catalyst used in the polymerization of the resin base of an ion-exchange resin can also assist in its depolymerization. Gas and vapor transport of nonradioactive organic species represents the preferred embodiment of this invention. However, this preference does not exclude other techniques such as solvent extraction of a nonradioactive organic phase away from an aqueous phase or a precipitated solid. (11) Another object of this invention is using a metallic hydroxide at least in part as the material which comprises metallic oxide. Complications would arise, for example, through formation of noxious gases arising from incineration of the inorganic groups which are attached to the resins to convert them to ion-exchange resins. This matter is discussed elsewhere--as noted, incineration of the noxious gases might require scrubbers which added to the radioactive volume actually required for waste disposal. As discussed elsewhere, removal of the inorganic species added into the original ion-exchange resin destroys the ion-exchange characteristics and their associated ability to hold water. Depolymerization also avoids some water retention by the resin. As noted earlier commercial practice demands that the ion-exchange resin cannot be disposed of dry because of the potential to expand and break its drums. By altering the solubility characteristics of the ion-exchange resin in organic solvents and water, the chemical changes imposed on the ion-exchange resin make feasible otherwise impractical separations processes such as aqueous-organic solvent extraction. For example, depolymerizing the ion-exchange resin may either directly liquefy the material produced or may transform the resin enough so it will dissolve more readily in a solvent. Here the liquid fluidity allows intimate contacting between phases in a way which is not feasible with solids. In this preferred embodiment, most material separated from radioactivity by vaporization is trapped in liquid form. This material can be monitored much more accurately than, for example, flowing gas. Water present in the hydroxide is used in steam distillation to assist the organic vapor transport. Water may be usefully added to the hydroxide to resupply a steam source. Likewise, other gases can be used as carrier gases for the organic vapor transport. And lowering the system pressure in a hermetically sealed condensation container can increase the boiling and improve the vapor transport over what would be met at higher pressure. Still other objects, advantages, and novel features of this invention will be apparent to those of ordinary skill in the art upon examination of the follow in a detailed description of preferred embodiments of the invention and the accompanying drawings. |
description | The present invention relates to control of fission reactions in a molten salt fission reactor. In particular, it relates to reversible methods of controlling the rate of neutron absorption in the reactor. Molten salt nuclear fission reactors are those where the fissile material is present in the form of a molten halide salt, usually chloride or fluoride. A novel design of such reactors was described in GB 2508537, in which the molten fuel salt was held in tubes surrounded by a second molten salt acting as a coolant. Control of reactivity of such reactors was proposed to be by using the negative temperature coefficient of reactivity to allow high temperatures to render the rector subcritical, by use of neutron absorbing control rods or by addition of the neutron absorbing material europium fluoride or cadmium fluoride to the coolant salt. Both europium fluoride and cadmium fluoride have severe limitations for use as neutron poisons added to the coolant salt. Europium is a strongly electronegative metal which would make reduction of the fluoride to the metal, either electrolytically or chemically, impossible without also reducing less electronegative coolant salt components such as zirconium. Cadmium is relatively easy to reduce to the metal, as set out in GB 2508537, but the metal produced is highly volatile and toxic at the temperature of the coolant salt and would therefore require specialised handling which, in the context of a nuclear reactor, would be complex and expensive. Europium and cadmium also have substantially lower absorption cross sections for fast neutrons than the boron more conventionally used as a neutron poison making them less useful. Conventional water cooled and moderated reactors use sodium borate added to the water to reduce reactivity, with the advantage that the borate is easily removed from the water as needed. There would be great advantage to having an analogous method available for the molten salt reactor described in GB 2508537, however borate salts are not chemically compatible with the molten salt of the coolant. Use of boron as the control material is particularly valuable in fast spectrum reactors as the boron has a high neutron absorption cross section even in the fast neutron spectrum. According to an aspect of the present invention, there is provided a method of controlling the reactivity of a molten salt fission reactor. The molten salt fission reactor comprises a core and a coolant tank, the core comprising fuel tubes containing a molten salt fissile fuel, and the coolant tank containing a molten salt coolant, wherein the fuel tubes are immersed in the coolant tank. The method comprises dissolving a neutron absorbing compound in the molten salt coolant, the neutron absorbing compound comprising a halogen and a neutron absorbing element. The method further comprises reducing the neutron absorbing compound to a salt of the halogen and an insoluble substance comprising the neutron absorbing element, the halogen being fluorine or chlorine, wherein the insoluble substance is not volatile at a temperature of the coolant during operation of the reactor. According to a further aspect, there is provided a method of controlling the reactivity of a molten salt fission reactor. The molten salt fission reactor comprises a core and a coolant tank, the core comprising fuel tubes containing a molten salt fissile fuel, and the coolant tank containing a molten salt coolant, wherein the fuel tubes are immersed in the coolant tank. The method comprises dissolving one or more neutron absorbing compounds in the molten salt coolant, wherein the one or more neutron absorbing compounds are chosen such that reduction of the neutron absorbing capacity of the one or more neutron absorbing compounds due to absorption of neutrons compensates for a fall in reactivity of the core in order to control fission rates in the core. According to a yet further aspect, there is provided apparatus for use in a nuclear fission reactor. The apparatus comprises an inlet, a mixing chamber, a filtration unit, and outlet, and a pump. The inlet is configured to be immersed in a pool of coolant salt of the nuclear fission reactor. The mixing chamber is configured to mix coolant drawn through the inlet with a reducing agent in order to reduce a neutron absorbing compound within the coolant salt into an insoluble substance containing a neutron absorbing element of the neutron absorbing compound, and a salt. The filtration unit is configured to filter the insoluble substance from the coolant salt. The outlet is configured to return the filtered coolant salt to the pool of coolant salt. The pump is configured to cause a flow of coolant salt from the pool through the outlet, then into the mixing chamber, then into the filtration unit, then out of the outlet. According to a yet further aspect, there is provided apparatus configured to operate in a nuclear fission reactor. The apparatus comprises an anode, a cathode, and a voltage regulator. The anode and cathode, are each configured to be immersed in a coolant salt of the nuclear fission reactor. The voltage regulator is configured to supply a voltage between the anode and cathode sufficient to electrolyse a neutron absorbing compound of the coolant salt and insufficient to electrolyse other components of the coolant salt. According to a yet further aspect, there is provided a nuclear fission reactor. The nuclear fission reactor comprises a core, a coolant tank, a neutron absorber addition unit, and a reduction unit. The core comprises fuel tubes containing a molten salt fissile fuel. The coolant tank contains a molten salt coolant and the fuel tubes are immersed in the coolant. The neutron absorber addition unit is configured to dissolve a neutron absorbing compound in the molten salt coolant, the neutron absorbing compound comprising a halogen and a neutron absorbing element. The reduction unit is configured to reduce the neutron absorbing compound to a salt of the halogen and an insoluble substance comprising the neutron absorbing element, the halogen being fluorine or chlorine, wherein the insoluble substance is not volatile at a temperature of the coolant during operation of the reactor. Further aspects and preferred features are defined in the dependent claims. It should be noted that, as used in this document, the terms below take the following meanings, which are standard within the chemical field: Element: A single type of atom (when used to describe a component of a compound) or a chemical formed from a single type of atom (when used to describe a substance). Compound: A chemical comprising two or more different elements which are bonded together by electrical forces. Substance: A chemical which cannot be separated into components by physical separation (i.e. without breaking chemical bonds), e.g. a compound, alloy or elemental substance, but not including a mixture. Reduction of reactivity of the nuclear reaction can be either for the purpose of temporary control of reactivity to compensate, for example, for an initially high reactivity which is expected to fall as fission proceeds and for full shut down of reactivity. This can be achieved by addition of a neutron absorbing material to the coolant salt. The material must be soluble in the coolant salt. Where temporary control of reactivity is desired, for example acting as a so called “reactivity shim” then a neutron absorbing material whose neutron absorbing properties are reduced after absorbing neutrons can be used so that the neutron absorber is progressively destroyed. Such absorbers are often referred to as burnable poisons Where longer lasting neutron poisons are required then slow burning poisons where the product of neutron absorption by the poison is also a neutron poison can be used. In the reactor described in GB 2508537 A, hafnium contamination of zirconium fluoride in the coolant salt acted as a slow burning poison. The amounts of burnable poisons added can be adjusted in order to ensure that the reduction in neutron absorption by the poisons compensates for the gradual fall in reactivity of the core, mitigating or removing the need for additional reactive material or additional neutron poisons to be added during the lifecycle of the reactor While use of cadmium and europium salts in this way is effective in shutting down the reactor, their use in non-emergency situations is problematical. Europium and hafnium are very reactive metals and cannot readily be removed from the coolant salt by reduction without also reducing the major salts in the coolant, such as zirconium, to their metallic forms. Cadmium is more easily reduced to the metal but, at the temperatures of the coolant salt, metallic cadmium is highly volatile and toxic and would therefore require specialised management which would be challenging in the context of a nuclear reactor. There is thus a need for more practical ways to control the reactivity of the reactor through addition of material to the coolant, where subsequent removal of the neutron poison from the coolant is practical. Two groups of chemicals have been found to be able to do this. First is the group of halides of relatively non reactive metals (Pauling electronegativity >1.5) which can be easily reduced to the metallic form either chemically or electrochemically, which have strong neutron absorption, and which form metals which are solids or non volatile liquids at the temperatures of the reactor. Indium and silver are also useful halides for this purpose, indium being liquid at the temperature of the coolant salt but having very low vapour pressure. Removal of the neutron poison is by reduction to the metallic form. For a coolant containing zirconium fluoride this is readily achieved by adding zirconium metal or zirconium difluoride to the coolant salt. For a coolant salt contain thorium tetrafluoride it can be done by adding thorium metal. Other reducing agents can also be used including the reactive group 1 and group 2 metals. The precipitated metal can then be removed by decantation or filtration or can be allow to fall to the bottom of the coolant tank and left indefinitely. An alternative to chemical reduction of the metal salt is electrochemical reduction where the metal produced can be accumulated as a deposit on the electrode or allowed to accumulate in contact with or in a container below the electrode where the metal is molten at the temperature of the reactor, as is the case with indium. The second group of chemicals are the sodium (or other group 1 metal) tetrafluoroborates. Sodium tetrafluoroborate is readily soluble in most molten salts. In the case of zirconium fluoride based molten salts it can be produced in situ by addition of borax to the salt which reacts to produce sodium tetrafluoroborate and zirconium oxide. Addition of a reactive metal such as zirconium or thorium to the coolant salt precipitates the boron in the form of a boride such as zirconium boride or thorium boride. While any metal with sufficient reactivity can be used, including strongly reducing metals such as sodium and potassium and metals of intermediate reactivity such as magnesium and calcium and metals of lower reactivity such as yttrium, scandium, zirconium, titanium and vanadium, metals whose fluorides are already components of the coolant salt such as zirconium or thorium have the advantage of not substantially changing the coolant salt composition. There is a further advantage if the metal added has a reactive lower valence halide, such as zirconium, vanadium and titanium di or trihalides which can either be generated in situ in the coolant salt or added directly instead of the metal. In this instance the reaction producing the metal boride is a solution reaction that proceeds rapidly to completion rather than a heterogeneous reaction between a solid and a liquid which can produce a layer of boride on the solid surface which inhibits further reaction. The precipitated boride can then be removed by filtration, decantation or other physical process. A particularly useful variant of these processes is to electrochemically reduce the coolant salt so that a boride based on a suitable metal halide present in the coolant salt such as zirconium or thorium tetrafluoride is formed as a layer on the electrode used. Removal of the precipitated boride is hence simplified and very precise control of the rate of boron removal can be achieved by controlling the electrochemical current density. Reversal of the electrochemical current may also be used to return boron to the coolant salt in a soluble state. FIGS. 1 and 2 each show a cross section of a reactor according to respective embodiments of the invention. In both figures, the reactor comprises a tank 101 containing coolant salt 102. Fuel tubes 103 are located within the coolant salt, forming the core of the reactor. Heat exchangers 104 withdraw the heat from the coolant salt, and flow baffles 105 are placed to improve convection of the coolant salt. The reactor of FIG. 1 additionally comprises a reduction unit comprising an electrochemical mechanism for removing a neutron absorber as described above. The electrochemical mechanism comprises a voltage regulator 1001 which may be located outside of the coolant salt, and an anode 1002 and cathode 1003 which are located within the coolant salt. The anode 1002 and cathode 1003 may be connected stacks of rectangular plates interleaved with one another with coolant salt between them. The electrode assembly (i.e. the anode and cathode) may be mounted inside the flow baffle structure in the reactor tank with electrical connections to the voltage regulator 1001. A reducible neutron absorbing compound as described above may then be removed from the coolant salt by applying a voltage across the anode 1002 and cathode 1003, causing electrolysis of the neutron absorbing compound and causing an insoluble substance containing the neutron absorbing element of the neutron absorbing compound to form on the cathode or anode (depending on the chemistry of the neutron absorbing compound). The insoluble substance containing the neutron absorbing element deposits on the relevant electrode and the neutron absorbing compound can be returned to the coolant salt by reversal of the current (reversing the electrolysis), alternatively, the electrode may be extracted and cleaned, and the neutron absorbing compound may be added separately. The flow baffle is perforated to allow mixing of the coolant salt inside the baffle to that in the remainder of the tank. The reactor of FIG. 2 comprises a mechanism reduction unit 2000 for removal of a neutron absorbing compound as described above by chemical reduction. The mechanism comprises an intake tube 2003, a pump 2001a and reducing agent mixing apparatus chamber 2001b, a filter filtration unit 2002, and a return tube 2004. Coolant salt containing the neutron absorbing compound is drawn up through the intake tube 2003 by the pump, and mixed with a reducing agent in the reducing agent mixing apparatus 2001. The reducing agent is selected in order to reduce the neutron absorbing compound into an insoluble substance comprising the neutron absorbing element and a salt comprising the non-neutron absorbing elements. The insoluble substance is then filtered out in the filter 2002. The filter 2002 may be, for example, a centrifugal filter. Alternatively, the filter 2002 may be a sedimentation tank, or any other suitable filter depending on the properties of the insoluble substance. The filtered coolant salt is then returned to the reactor tank via the return tube 2004. The reactors of FIG. 1 and FIG. 2 may comprise neutron absorber addition units 1100, 2100 configured to dissolve the neutron absorbing compound in the molten salt coolant. |
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050769715 | description | An apparatus for use in showing how various radioactive materials can be decontaminated in accordance with the teachings of the present invention is broadly denoted by the numeral 10 and is shown in FIG. 1. The purpose of the present invention is to stimulate charged particles inside the atomic system of a radioactive material thereby rapidly accelerating the rate of decay of alpha, beta and gamma particles from the material and thereby decontaminating the material. The decontamination is accomplished in the apparatus of FIG. 1 by the application of a stimulus in the form of a negative electrical charge potential in close proximity to the nucleus of a sample of radioactive material. A large negative potential has the effect of lowering the energy barrier which retains the positively charged particles, such as alpha particles, within the nucleus. As the negative charge potential is placed upon the atomic nuclei, the rate of emission of alpha, beta and gamma particles is increased to thereby accelerate decontamination of the radioactive materials. Generator 10 includes sphere 13 forming part of a generator mechanism 14. A radioactive sample 15 to be enhanced or decontaminated is placed on a platform 16 supported by a bracket 17 on the interior of sphere 13 near the upper end thereof adjacent to a hole 19 in the sphere. Thus, the radioactive sample is within the sphere and will be subjected to the electrostatic potential generated in the sphere as hereinafter described. The sphere 13 is supported on legs 22 on a base 23 which is grounded. Thus, the sphere 13 and sample 15 are isolated from external electrical fields. For generator 10 to operate, sphere 13 must be maintained in spatial and electrical isolation from all other elements including the base plate 23. To this end, legs 22 must be electrical insulators. Sphere 13 receives electrical charges by way of an insulated moving belt 24 which extends between an interior pulley 26 within sphere 13 and an exterior pulley 28 carried in some suitable manner and on base 23. Drive mechanism 30 is a motor providing rotation of motion to exterior pulley 128. A high voltage generator is located near pulley 28 near the base plate 23. Generator 36 delivers charges to belt 24 by way of a pair of electrically conducting needles 38 which contact belt 24 on either side of pulley 28. Generator 26 is typically capable of delivering voltages in the range of 50,000 to 500,000 volts. The purpose of generator 10 is to provide a large negative electrostatic potential with no field at the site of the sample 15. This can be accomplished by placing the sample 15 anywhere within or on the sphere 13. The radioactive sample 15 can comprise an alpha, beta or gamma emitter. An alpha emitter defining the sample 15 can be, for instance, thorium 230, uranium 235 or plutonium 239. These three sources have half lives of 8.times.10.sup.4 years, 7.1.times.10.sup.8 years and 24,360 years, respectively. There are a few hundred alpha emitters with half lives ranging from less than a millisecond (Fr 215) to billions of years (uranium 238). Tests were conducted with generator 10 with sample 15 located as shown in FIG. 1. These tests were conducted with the use of a Geiger-Meuller tube 40 adjustably carried by a tube 42 secured by an annular ring 44 to the outer surfaces sphere 13. Tube 42 surrounds hole 19 so that alpha, beta or gamma particles emitted from sample 15 will be directed to tube 40 and sensed thereby. A scalar 46 is coupled by a cable 48 to Geiger-Mueller tube 40. In the experimental work, three radioactive sources were used as sample 15. The principal source was thorium 230 with an activity of 0.1 ci. As thorium oxide, the sample was electrodeposited and diffusion bonded on platform 17 which, for purposes of illustration, was a 0.001 inch platinum plate in a metal cylinder with a diameter of 24 millimeters and a thickness of three millimeters. This source was made to specification by the Isotope Products Laboratories, of Burbank, Calif. The other sources included a sample of pitchblende obtained from Ward's Natural Science Establishment, and cesium 137 in a cylindrical plastic holder from Nucleus, Inc. of Oak Ridge, Tenn. The Geiger-Meuller tube 40 and scalar 46 were obtained from Nucleus, Inc. of Oak Ridge, Tenn. The Geiger-Mueller tube (model PK2) detects alpha, beta and gamma particles. The scalar 46 (model 500) was coupled by cable 48 to the Geiger-Mueller tube, the cable being an eight foot coaxial MHP cable to shield the same against the effects of the high voltage generator 10. Generator 10 was a 250,000 volt generator of negative polarity. It was obtained from Wabash Instrument Company (model N 100-V) of Wabash, Ind. The diameter of sphere 13 of the generator was approximately 25 centimeters. The sample 15 was housed in a metal clamp inside the sphere 13. This clamp was annular base 44 which can be wood or plastic on the outside of sphere 13. The sensor tube 40 was inserted to various depths into a 3.5 centimeter diameter hole in base 44. The size of hole 19 was 15 millimeters at the top of sphere The voltage achieved with the particular Van de Graaff generator was approximately 50,000 volts. Measurements of the voltage were made from spark lengths by estimating 25,000 volts per inch. A better measure is provided by the source-to-sensor distance. This gives reasonable voltage values if the speed of belt 24 is increased slowly to the point where there is an electron discharge and the scalar goes off scale. The present invention postulates that an external, electrostatic potential penetrates the interior of the nuclei of a radioactive material to the nuclear well. The material should be an electrical conductor and be housed in a metallic environment. The generator is a simple and convenient high voltage source which acts as a stimulus for accelerated. On the spherical surface of radius a, the voltage is equal to Q/a, where Q is the charge, negative or positive, delivered by the belt. This potential is constant inside the sphere 13 where the electrical field is zero at all locations within the sphere. A series of experiments were carried out with thorium 230, and the experiments proved to be successful in that a substantial change in activity occurred when the generator 10 was switched from an off condition to an on condition. Over 300 experimental readings were taken which exhibit positive or negative enhancement. Qualitatively, the measurements always agreed with the theory. Table 1 shows, for thorium 230, theoretical and measured values of enhancement versus the potential of the generator 10. FIG. 2 shows a plot of the values set forth in Table 1, and the straight line in FIG. 2 is theoretical value, the data points showing the agreement between the theoretical and experimental values within experimental errors. The enhancement values have a standard deviation of about five percent. Each point on the graph is represented by about 20 readings. The voltage reading are accurate to 1.8 kv. The principal experimental difficulty was in measuring the voltage of the generator, Some of the values of 1.eta..lambda./.lambda..sub.o values were much too large. These values were attributed to errors in calculating the magnitude of the voltage. Such a value is shown by a data point is denoted by an asterisk in Table 1. TABLE 1 ______________________________________ Theoretical and measured values .epsilon. vs .phi. for Thorium 230 .phi. in kv .epsilon.th .epsilon.m % difference ______________________________________ Negative voltages -3.94 0.13 0.15 13.3 -9.37 0.34 0.42 25.0 -18.8 0.80 0.91 13.8 -21.9 0.99 1.13 14.1 -30.0 1.56 1.55 0.64 -33.0 1.81 1.80 0.52 -42.3 2.77 3.14 13.4 Positive voltages +13.3 -.341 -0.310 9.1 +14.8 -0.370 -0.652 76.2* +22.6 -0.508 -0.494 2.76 ______________________________________ The mineral pitchblende consists of about 70% uranium oxide and about 7% thorium oxide with lesser amounts of several stable oxides. Natural uranium is primarily uranium 238. At two generator speed settings, the activity increased appreciably as was expected. At .phi.=-22.6kv, .lambda./.lambda..sub.o equals 1.97.+-.0.37. Within experimental error, this agrees with the theoretical value of 2.35. The large range for the measured .lambda./.lambda..sub.o is due to the fact that the activity at .phi.=0 was only 2.23 times the background count. Cesium 137 decays by beta emission to Ba 137, which is stable with a half-life of 30.2 years. A change in .lambda. was detected as the applied voltage of the generator 10 was turned on. The magnitude of the effect was much smaller. The foregoing description relates to the decontamination or enhancement of the decay rate of a radioactive material. A typical potential or voltage value for such enhancement is in the range of 40 to 50 kilovolts and a typical radioactive material suitable for showing enhancement is thorium 230. The ignition can be accomplished by a Van de Graaff generator 10 in which the radioactive source 15 is within or on the sphere 13 of the generator. A typical voltage is 350 kilovolts, and the ignition time is typically one hour. An initial ignition voltage of about 300 kilovolts for a period one hour may well be sufficient for igniting a nuclear fuel rod in the sphere of the generator. If necessary, a second ignition step may be used to complete the decontamination process. The mechanism for alpha depletion differs from the mechanisms for beta and gamma depletion which are slower. In alpha depletion, the Coulomb barrier is 2Z.sub.1 e.sup.2 /r is modified by a constant term that is: 2Z.sub.1 e.sup.2 /r-2e.phi.. Variations are present but they are not as significant as the constant term. Here .phi. is the applied voltage on the generator terminal. ln.lambda./.lambda..sub.o =3.71Z.sub.1 (1/E.sup.1/2 -1/(E+2e.phi.).sup.1/2). Here p80 equals the decay rate and .lambda..sub.o equals the quiescent decay rate. Z.sub.1 is the charge of the daughter nucleus and E is equal to the alpha decay energy. The mechanism for beta decay involves contact between the electrons and the nucleus. This is a short range not a long range interaction. In the decay of thorium 234, the electrons which make contact with the nucleus are the S electrons. They have zero angular momentum. Thorium has the same number of S electrons as uranium, namely 14. In thorium, there are 76 electrons in the g, d and f angular momentum states. They do not contribute as much to the beta decay as do S electrons. The half-life is 24 days. To achieve ignition of the radioactive materials, all that is needed is some mechanism to excite the charged particles. The following technique is suitable: 1. Place a sample in contact with a Van de Graaff generator operating at a modest voltage for 10 or 15 minutes. On large samples the Van de Graaff generator is a most effective source for establishing the ignition. It establishes a voltage throughout the entire sample. Gamma decay enhancement, like alpha decay enhancement, is long range but there is no Coulomb barrier to magnify the effect. All nuclei change their shapes from spherical to ellipsoidal etc. Gamma radiation occurs as a result of the oscillations of the protons and neutrons in the nucleus. Tests were conducted to show that a positive or negative voltage on a Van de Graaff generator accelerates beta and alpha decay. One beta and two alpha emitters were placed inside the generator sphere, charged to a voltage of 350+75 kv, for a period of twelve hours. When the voltage was switched off, the measured activity oscillated through substantial variations. After three days the measured depletion was about 1% for Tl 204, about 7% for Po 210 and about 2.6% for Th 230. After seven days, the depletion had increased to about 5.3%, about 55.3% and about 81.8%, respectively. It is expected that the depletion will continue to background for all three sources within about 60 days. A depletion "burn" can be initiated in an alpha emitter with a Van de Graaff voltage of about -50 kv in a time interval of 20 minutes or so. The alpha depletion is primarily due to the alpha excitation 2e.phi.. The test procedure was as follows: Three radioactive sources: Tl 204, a beta emitter and Th 230 and Po 210, both alpha emitters, were put inside the terminal of a Van de Graaff generator. The voltage was left on for 12 hours of consecutive running time. The quiescent activity A.sub.o of each source was measured before insertion. Shortly after the generator was turned off, the activity was monitored with a Geiger-Meuller counter. All three samples exhibited oscillations in the counting rate similar to that of a weakly damped harmonic oscillator. The oscillations continued for more than two weeks, indicating that the new quiescent value of A.sub.1 was close to background. The generator was operated at about 3/4 maximum speed. The generator was kept away from nearby conductors, which might draw off the charge. The voltage was measured by observing the spark gap distance. These varied from as low as 6 inches (150 kv). The average terminal voltage was estimated to have been (350.+-.75) Kv. In ln Coulomb barrier modification, a voltage of 412.5 kv is much more effective in enhancing alpha decay than a voltage whose magnitude is 62.5 kv less. This is because .lambda./.lambda..sub.o depends on .phi. exponentially. Tl 204 decays by beta minus emission to Pb 204, a stable isotope, with a half-life of 3.8 years. The corresponding decay rate .lambda.=5.78.times.10.sup.-9 sec.sup.-1. The quiescent depletion of Tl 204 in a period of seven days is EQU D.sub.o =.lambda..sub.o t=0.349% Measured values for A.sub.o and A after seven days were found to be EQU A.sub.o =673.9.+-.0.11c/s and EQU A=638.0.+-.4.2c/s The depletion or decontamination at this time was EQU D=(A.sub.o -A)/A.sub.o =5.33% This is 15 times D.sub.o. Three hours and 30 minutes later the measured activity, A, was 4.47% higher than A.sub.o. The Tl 204 sample was provided by the Nucleus Inc., Oak Ridge, Tenn. It was housed in a plastic holder. In the theory of beta decay the rate of decay is proportional to the electron charge density at the nucleus .rho.(o)=e.psi.*.psi.(o). A negative voltage .phi. decreases the potential energy of th atomic electron and vice versa. This displaces the electron cloud away from the nucleus, increasing .rho.(o). During the operation of the Van de Graaff, with the source inside the terminal, .rho.(o,.phi.) has a steady state value. Polonium 210 decays by alpha emission to Pb 206, a stable isotope with a half-life of 138.4 days. The corresponding decay rate if .lambda..sub.0 =5.80.times.10.sup.-8 /sec. The decay energy is E=5.40 MeV. The quiescent depletion of Po 210 in seven days is EQU D.sub.o =.lambda.t=3.51% The measured values for A.sub.o and A after seven says were EQU A.sub.o =332.13.+-.1.52c/s The depletion at this time was EQU D=(A.sub.o -A)/A.sub.o =53.86% This is about 15.3 times D.sub.o. Twelve hours later the measured activity A was 200 c/s, 30% lower than A.sub.o. The oscillating period for this sample of A.sub.o is about one day. The alpha depletion studies on Po 210 indicate that there is one significant mechanism which modifies the Coulomb barrier. This effect is described by Eq. (5) above where V.sub.a represents an increase in the alpha particles potential energy when the Van de Graaff voltage .phi. is negative and vice versa. Thorium 230 decays by alpha emission to Ra 226 with a decay energy of 4.767 MeV. The half-life is 80,000 years. There are about a dozen daughters in the Th 230 decay scheme. The first daughter Ta 226 is an alpha emitter with a half-life of 1,600 years. The successive daughters are short half-life alphas and betas. The chain proceeds to Pb 210, which decays by alpha and beta emission with a half-life of 21 years. Subsequent daughters lead to Po.sup.210 and then to Pb 206, a stable isotope. The quiescent decay constant for Th 230 is EQU .lambda..sub.o =2.75.times.10.sup.-13 /sec The quiescent depletion in seven days is: EQU D.sub.o =.lambda.t=1.66.times.10.sup.-5 T Our measured values for A.sub.o and A.sub.1, after days, were EQU A.sub.a =91.47.+-.4.57c/s EQU A=16.85.+-.0.04c/s The depletion at this time was EQU D=(A.sub.o -A)/A.sub.o =81.22% This is 4.89.times.10.sup.4 times greater than D.sub.o. Sixteen hours later the measured activity A=24.64 c/s, an increase of 46% over our earlier low count, but substantially less than A.sub.o. The fact that the depletion rate is much faster in Po and Th than in Tl is understandable. The beta decay process involves electron-nuclear contact e.psi.*.psi.(o) which is measured by the steady state and transient behavior of the atomic electron cloud. The alpha decay process is controlled by the Coulomb barrier, as modified. A small change in the charged density of the atomic electrons has a magnified effect on the decay rate. The Van de Graaff voltage .phi. ignites radioactive waste. If the burn is going too slowly, re-ignite with an e.phi..DELTA.t less than the initial value. High voltages may be hazardous. For example. .phi.=2 MV predicted to convert the half-life of U.sup.238 to one second. Before initiating a decontamination procedure, the composition of the fuel should be determined. ______________________________________ DECAY STEP HALF LIFE (t.sub.1/2) ______________________________________ DECAY OF CHAIN OF URANIUM 235 (1) .sub.92 U.sup.235 .fwdarw. .sub.90 Th.sup.231 + .alpha. 7.13 .times. 10.sup.8 years (2) .sub.90 Th.sup.231 .fwdarw. .sub.91 Pa.sup.231 + .beta. - 25.6 hours (3) .sub.91 PA.sup.231 .fwdarw. .sub.89 Ac.sup.227 + .alpha. 3.25 .times. 10.sup.4 years (4) .sub.89 Ac.sup.227 .fwdarw. .sub.90 Th.sup.227 + .beta. - 21.6 years (5) .sub.90 Th.sup.227 .fwdarw. .sub.88 Ra.sup.223 + .alpha. 18.5 days (6) .sub.88 Ra.sup.223 .fwdarw. .sub.86 Rn.sup.219 + .alpha. 11.43 days (7) .sub.86 Rn.sup.219 .fwdarw. .sub.84 Po.sup.215 + .alpha. 4.0 seconds (8) .sub.84 Po.sup.215 .fwdarw. .sub.82 Pb.sup.211 + .alpha. 1.78 .times. 10.sup.- 3 seconds (9) .sub.82 Pb.sup.211 .fwdarw. .sub.83 Bi.sup.211 + .beta. - 36.1 minutes (10) .sub.83 Bi.sup.211 .fwdarw. .sub.81 Tl.sup.207 + .alpha. 2.15 minutes (11) .sub.81 Tl.sup.207 .fwdarw. .sub.82 Pb.sup.207 + .beta. - 4.78 minutes .sub.82 PB.sup.207 is stable. DECAY OF CHAIN OF URANIUM 238 (1) .sub.92 U.sup.238 .fwdarw. .sub.90 Th.sup.234 + .alpha. 4.51 .times. 10.sup.9 years (2) .sub.90 Th.sup.234 .fwdarw. .sub.91 Pa.sup.234 + .beta. - 24.1 days (3) .sub.91 PA.sup.234 .fwdarw. .sub.92 U.sup.234 + .beta. - 6.66 hours (4) .sub.92 U.sup.234 .fwdarw. .sub.90 Th.sup.230 + .alpha. 2.48 .times. 10.sup.5 years (5) .sub.90 Th.sup.230 .fwdarw. .sub.88 Ra.sup.226 + .alpha. 80.0 years (6) .sub.88 Ra.sup. 226 .fwdarw. .sub.86 Rn.sup.222 + .alpha. 1622 years (7) .sub.86 Rn.sup.222 .fwdarw. .sub.84 Po.sup.218 + .alpha. 3.823 days (8) .sub.84 Po.sup.218 .fwdarw. .sub.82 Pb.sup.214 + .alpha. 3.05 minutes (9) .sub.82 Pb.sup.214 .fwdarw. .sub.83 Bi.sup.214 + .beta. - 26.8 minutes (10) .sub.83 Bi.sup.214 .fwdarw. .sub.84 Po.sup.214 + .alpha. 19.7 minutes (11) .sub.84 Po.sup.214 .fwdarw. .sub.82 Bi.sup.210 + .beta. - 164 seconds (12) .sub.82 Pb.sup.210 .fwdarw. .sub.83 Bi.sup.210 + .beta. - 21 years (13) .sub.83 Bi.sup.210 .fwdarw. .sub.84 Po.sup.210 + .beta. - 5.0 days (14) .sub.84 Po.sup.210 .fwdarw. .sub.82 Pb.sup.206 + .alpha. 138.4 days .sub.82 Pb.sup.206 is stable DECAY CHAIN OF PLUTONIUM 239 (1) .sub.94 Pu.sup. 239 .sub.92 U.sup.235 + .alpha. 24,360 years Then follow decay chain for Uranium 235. ______________________________________ |
abstract | Nuclear fuel assemblies include non-symmetrical fuel elements with reduced lateral dimensions on their outer lateral sides that facilitate fitting the fuel assembly into the predefined envelope size and guide tube position and pattern of a conventional nuclear reactor. Nuclear fuel assemblies alternatively comprise a mixed grid pattern that positions generally similar fuel elements in a compact arrangement that facilitates fitting of the assembly into the conventional nuclear reactor. |
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054460751 | claims | 1. Apparatus for providing manipulative physical therapy to a patient, comprising: a first mass of putty including a reaction product of a polysiloxane and either a boron- or a tin-containing compound and having a first color, and further including an unreacted, uncured second polydiorganosiloxane gum; and at least one additional mass of putty including (a) a reaction product of polysiloxane and either a boron- or tin-containing compound, (b) an unreacted, uncured polydiorganosiloxane gum, or (c) mixtures of (a) and (b), said additional mass having a second color distinct from the first color, the additional mass adaptable to be manually combined by the patient with the first mass to form a combined mass having a uniform color which is a result of blending the first color and the second colors. 100 parts by weight of a chain-extended polysiloxane reaction product formed by reacting a polysiloxane having a viscosity of 50,000 to 100,000 centistokes with a reactant containing oxygen and either boron or tin; 10 to 50 parts by weight of an unreacted, uncured second polysiloxane gum having a Williams' plasticity in the range of 120 to 140 mm; 0.2 to 2.0 parts of an internal lubricant; and 5 to 45 parts by weight of a particulate filler material. supplying to a patient a first malleable mass of putty comprising (a) a chain-extending reaction product of polydiorganosiloxane and a compound containing boron or tin, and (b) an unreacted, uncured second polydiorganosiloxane gum, said first mass having a first color; supplying to the patient a second malleable mass of putty having a second color different from the first color, the second malleable mass being blendable into the first mass; directing the patient to combine the first mass with the second mass to form a combined mass having a uniform color which is a result of blending the first color with the second color, the uniform color of the combined mass indicating that the patient has manipulated the combined mass by a predetermined amount. after a uniform color has been achieved in the combined mass, supplying to the patient a third malleable mass having a third color which is different from the uniform color of the combined mass; directing the patient to manually combine the third mass with the combined mass to form a second combined mass having a second uniform color which is a result of blending the third color with the uniform color, the achievement of the second uniform color of the second combined mass indicating that the patient has again manipulated the combined mass by a predetermined amount. for a predetermined number of iterations, performing the following steps: after a uniform color has again been achieved in the combined mass, supplying to the patient a further malleable mass having a color different from the uniform color; and directing the patient to manually combine the further malleable mass with the combined mass until a uniform color is again achieved. 2. The apparatus of claim 1, wherein the additional mass is significantly smaller and more highly colored than the first mass. 3. The apparatus of claim 1, and further comprising a plurality of additional masses having colors distinct from the first color, at least some of the colors of the additional masses being distinct from each other, a first additional mass adaptable to be combined with said first mass to form a combined mass, remaining additional masses adapted to be successively manually added to the combined mass, respective uniform colors being achieved after the complete blending of each additional mass with the combined mass. 4. The apparatus of claim 1, wherein said first mass comprises: 5. The apparatus of claim 4, wherein said chain-extended siloxane reaction product is selected from the group consisting of borosiloxane and stannosiloxane. 6. The apparatus of claim 5, wherein said reactant is dibutyldiacetoxytin. 7. The apparatus of claim 5, wherein said reactant is trimethyl boroxine. 8. The apparatus of claim 4, wherein said particulate material is precipitated silica. 9. A method of manipulative therapy, comprising the steps of: 10. The method of claim 9, wherein the second malleable mass is significantly smaller and has a more concentrated color than the first malleable mass. 11. The method of claim 9, and further comprising the step of inspecting the combined mass for streaks, the presence of streaks indicating that the combined mass has been manipulated by less than a predetermined amount. 12. The method of claim 9, wherein the first malleable mass is a putty including borosiloxane. 13. The method of claim 9, wherein the second malleable mass is a putty including a polymer constituent selected from the group consisting of borosiloxane, a polydiorganosiloxane gum and mixtures thereof. 14. The method of claim 9, and further comprising the steps of: 15. The method of claim 14, and further comprising: |
051630788 | claims | 1. A multilayer film reflecting mirror comprising layers of a plurality of substances constructed on a substrate, suited for X rays having wavelengths of 100 .ANG. and less, wherein a deviation .DELTA. from a standard value of thickness of each layer is within a range defined by ##EQU11## where .theta. is the grazing angle of an X ray incident on the reflecting mirror and .lambda. is the wavelength of the X ray. 2. The reflecting mirror according to claim 1, wherein said plurality of substances is comprised of Ni and Sc, and said layers are alternately laminated on the substrate so that a multilayer film is fabricated. 3. The reflecting mirror according to claim 1, wherein said plurality of substances is comprised of Re and Al, and said layers are alternately laminated on the substrate so that a multilayer film is fabricated. 4. The reflecting mirror according to claim 1, wherein said plurality of substances is comprised of W and C, and said layers are alternately laminated on the substrate so that a multilayer film is fabricated. 5. The reflecting mirror according to claim 1, wherein at least one of said plurality of substances is comprised of Ni, and said layers are alternately laminated on the substrate so that a multilayer film is fabricated. 6. The reflecting mirror according to claim 1, wherein said substrate has a curvature. 7. The reflecting mirror according to claim 1, wherein said reflecting mirror constitutes a Schwarzschild optical system. 8. The reflecting mirror according to claim 1, wherein said reflecting mirror constitutes a Walter optical system. |
042696615 | claims | 1. In a nuclear reactor fuel assembly including an array of fuel rods and control rod guide tubes held in spaced relationship with each other by means positioned at intervals along their length, upper and lower nozzles connected to opposite ends of said guide tubes, and an upper core plate above the fuel assembly which absorbs lifting forces acting on the assembly, the upper nozzle including: control rod guide tube extensions including a center guide tube extension and selected guide tube extensions, said extensions being attached to an orifice plate located immediately above said fuel rods; means respectively connecting the control rod guide tubes to said tube extensions; a hold-down plate vertically movable on said tube extensions, said plate having openings of a size sufficient to accept the guide tube extensions; springs on said selected tube extensions between the orifice and hold-down plate and biased in a direction to urge the hold-down plate vertically against said upper core plate; at least one vertically extending slot in each guide tube extension outer surface; and a pin for each of said slots, each of said pins being mounted in the hold-down plate and in a position to have its end project into a corresponding slot formed in the tube extension, the arrangement being such that as upwardly acting forces lift the fuel assembly, the springs on said extensions are unformly compressed until the pins are engaged by the bottom of said slots, thus permitting the assembly to be lifted to its upper limit without compressing the springs to a solid condition; and a shoulder on said center tube extension arranged to contact the hold-down plate when said forces lift the fuel assembly to thereby supplement the action of the pin and slot arrangement in limiting upward movement of the fuel assembly. |
summary | ||
047568720 | abstract | The invention relates to a nuclear power station for a gas cooled high temperature pebble bed nuclear reactor. The nuclear power station is characterized by a combination of features, whereby the system inherent properties of a high temperature reactor are utilized to make possible the economical operation of a nuclear power station of medium capacity (300-600 MW.sub.el) while maintaining a high standard of safety. The characteristics comprise a reactor protection building equipped with pressure relief means in combination with filters, several auxiliary cooling systems separate from the operating cooling systems for the removal of decay heat in the case of accidents, and the utilization of a liner cooling system for the prestressed concrete reactor pressure vessel to assure the removal of the decay heat in case of a failure of the auxiliary cooling systems. |
abstract | According to the present invention, an improved source collimator for use in nuclear medicine imaging is provided. The improved source collimator utilizes a larger collimation angle than has previously been used in the art. The use of the larger collimation angle for the source collimator reduces the sensitivity of the nuclear medical imaging systems to misalignment between the detector and the source collimators. |
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042723205 | claims | 1. In a system comprising means for extracting useful energy in a controlled manner from a target imploded by energy from at least one laser beam; the improvement comprising: a high density, substantially spherically configured, concentric shelled target consisting of a pusher shell consisting of Au, said pusher shell containing therein a quantity of fuel consisting of DT; an ablator-pusher shell consisting of SiO.sub.2, said ablator-pusher being positined in spaced relation about said pusher shell forming a region therebetween filled with a material consisting of plastic foam. 2. The target defined in claim 1, wherein said pusher shell has an inner radius of about 0.0057 cm and an outer radius of about 0.0068 cm, and wherein said ablator shell has an inner radius of about 0.0400 cm and an outer radius of about 0.0420 cm. 3. The target defined in claim 1, wherein said pusher shell has a density of about 19.3 gm/cm.sup.3 and a mass of about 10.46 .mu.gm, wherein said quantity of fuel has a density of about 0.05 gm/cm.sup.3 and a mass of about 0.0388 .mu.gm wherein said ablator shell has a density of about 2.5 gm/cm.sup.3 and a mass of about 105.6 .mu.gm; and wherein said plastic foam has a density of about 0.02 gm/cm.sup.3 and a mass of about 5.33 .mu.gm. |
description | The present invention relates to a radiation detector for detecting X-rays, gamma rays, and the like; and, more particularly, the invention relates to an oxide phosphor, that is suitable for use in a radiation detector such as an X-ray CT apparatus and a positron camera, and to a radiation detector and an X-ray CT apparatus in which an oxide phosphor is used. A radiation detector for an X-ray CT apparatus there has conventionally used a xenon gas chamber, a combination of bismuth germanium oxide (BGO single crystal) and a photomultiplier tube, as well as combinations of a Csl:Tl single crystal or CdWO4 single crystal and a photodiode. In recent years, a rare-earth phosphor having good radiation-to-light conversion efficiency has been developed, and a radiation detector using a combination of such a phosphor and a photodiode has already been put into practical use. A rare-earth phosphor consists of a rare-earth element oxide or a rare-earth element oxysulfide as a base material, to which is added an activator that serves as a luminescent component. As a rare-earth element oxide phosphor, a phosphor including yttrium oxide and gadolinium oxide as a base material, and a phosphor represented by a formula (Gd1-xCex)3Al5-yGayO12 have been proposed in Japanese Patent Laid-open Publication JP-A-3-50991 and in International Patent Laid-open Publication WO99/33934 (PCT/JP98/05806), respectively. Properties generally required of a scintillator material of the type used in a radiation detector include high luminescence efficiency, short afterglow, and high X-ray stopping power. There are phosphors having high luminescence efficiency among the above-mentioned phosphors, but the afterglow time thereof is relatively long. When the afterglow of the scintillator used in X-ray detectors of an X-ray CT apparatus is large, the acquired information becomes indistinct along the time-axis. Many conventional scintillator materials have a problem with large afterglow. However, the phosphor (Gd1-xCex)3Al5-yGayO12 (0.0005≦x≦0.02, 0<y<5), mentioned in the International Patent Laid-open Publication WO99/33934, has excellent scintillator properties, including both high light emission output and short afterglow. Although it has been confirmed that (Gd1-xCex)3Al5-yGayO12 material in powder form certainly has good scintillator properties, it has been also found that a new problem occurs when manufacturing a scintillator plate using this powder composition. That is, the luminescence efficiency and the afterglow greatly fluctuate when sintering the scintillator plate, and it is hard to obtain stable properties. Especially, it has been revealed that the afterglow properties are significantly deteriorated by sintering, and the thickness of the sintered body increases as well. A thick sintered body is indispensable for a low cost manufacturing technique, so that there has been a need for a technique for manufacturing a thick sintered body with low afterglow. An object of the present invention is to provide a phosphor which has a thick plate or block shape that is suitable for mass production, while also having a high luminescence efficiency to X-rays, an extremely low afterglow, and good reproducibility of these properties, by solving the instability of the scintillator properties, which is a drawback of the phosphor of (Gd1-xCex)3Al5-yGayO12 composition. Another object of the present invention is to obtain a radiation detector with large light emission output and low afterglow by using the above-described phosphor as a scintillator of a radiation detector having a light detector. Still, another object of the present invention is to provide a tomogram having high resolution and high quality by applying the above-descibed radiation detector to an X-ray CT apparatus. To achieve the above-described objects, three-component phosphors (sintered bodies) consisting of (Gd,Ce)—Al—Ga with various compositions were manufactured, and the afterglow thereof was measured. As a result, it was found that scintillator material having low afterglow with significantly greater stability can be obtained, regardless of the plate thickness of the sintered body or the shape and size of the sintered body, when the composition is in a predetermined non-stoichiometric region. Consequently, the invention has been based on this finding. That is, the present invention provides an oxide phosphor and a sintered body thereof, described below: (1) an oxide consisting at least of Gd, Ce, Al, Ga, and O, the crystal structure of which is a garnet structure, wherein the atomic ratio (Gd+Ce)/(Al+Ga+Gd+Ce) is more than 0.375 and no more than 0.44, and the atomic ratio Ce/(Ce+Gd) is no less than 0.0005 and no more than 0.02. (2) an oxide phosphor and a sintered body thereof according to paragraph (1), wherein the atomic ratio of Ga/(Al+Ga) is more than 0 and less than 1.0. (3) an oxide phosphor and a sintered body thereof according to paragraphs (1) or (2), having a perovskite structure as an impurity phase. (4) an oxide phosphor and a sintered body thereof according to paragraph (3), wherein the intensity of the main diffraction line of the perovskite structure in X-ray diffraction measurement is 50% or less of the intensity of the main diffraction line of the garnet structure. Further, the present invention provides a radiation detector having a ceramics scintillator and a light detector for detecting light emission of the scintillator, the ceramics scintillator using an oxide phosphor according to the foregoing paragraphs (1) to (4). Further, the present invention provides an X-ray CT apparatus having an X-ray source; an X-ray detector placed opposite to the X-ray source; a rotating circular plate supporting the X-ray source and the X-ray detector, which is driven to rotate around an object to be examined; and an image reconstructing means for image-reconstructing a tomogram of the object, based on the intensity of X-rays detected by the X-ray detector, wherein the above-described radiation detector is employed as the X-ray detector. Hereinafter, the details of the phosphor according to the present invention will be described. The phosphor of this invention is an oxide phosphor including Gd, Al, and Ga as base elements and Ce as the luminescence component, similar to the phosphor (Gd1-xCex)3Al5-yGayO12(Gd1-xCex) (0.0005≦x≦0.02, 0<y<5) described in the international publication WO99/33934 of the international patent application by the inventors of the present invention. While the one mentioned in the international publication WO99/33934 is a phosphor having a stoichiometric composition, the one according to the present invention has a non-stoichiometric composition. Incidentally, the description of constituent elements of the phosphor mentioned in the specification of the international publication WO99/33934 constitutes a part of this specification. FIG. 3 is a composition diagram of a (Gd,Ce)−Al—Ga three-component system. In the drawing, the line designated by reference character S represents the phosphor of the formula (Gd1-xCex)3Al5-yGayO12, described in the international publication WO99/33934, and according to the notation system in this invention, it is described by the relations (Gd+Ce)/(Al+Ga+Gd+Ce)=⅜=0.375, 0.0005≦Ce/(Ce+Gd)≦0.02, 0<Ga/(Al+Ga)<1. The phosphor according to the invention is the one in the composition region K of a non-stoichiometric composition, shown by hatching in FIG. 3. More specifically, the atomic ratio (Gd+Ce)/(Al+Ga+Gd+Ce) is more than 0.375 and no more than 0.44, and the atomic ratio of Ce/(Ce+Gd) is no less than 0.0005 and no more than 0.02. If the atomic ratio (Gd+Ce)/(Al+Ga+Gd+Ce) is 0.375 or less, among the scintillator properties, the afterglow particularly increases by a hundredfold, and the light emission output decreases by 30% or more. On the other hand, if the atomic ratio (Gd+Ce)/(Al+Ga+Gd+Ce) is above 0.44, the damping factor of the afterglow 300 ms after stopping the excitation source is a stably low value, for example, about 4×10−5. However, the Gd(Ga,Al)O3-phase of the perovskite structure constituting a hetero phase is generated, comprising 50% or more of the host crystal of the garnet structure. Further, the light emission output is deteriorated by 40%. If the ratio Ce/(Ce+Gd) is outside the above range, sufficient light emission cannot be obtained. The atomic ratio Ga/(Al+Ga) is preferably more than 0 and less than 1.0. If this atomic ratio is out of the above range, sufficient emission intensity cannot be obtained, even if Ce is doped. The above-described change of the scintillator properties caused by variation of the atomic ratio (Gd+Ce)/(Al+Ga+Gd+Ce) can be understood from the following consideration of the crystal structure. FIG. 4 shows a garnet structure having a theoretical stoichiometric composition represented by (Gd1-xCex)3Al5-yGayO12. The ion radii of Gd and Ce are 0.97 Å and 1.07 Å respectively, and those of Al and Ga are 0.51 Å and 0.62 Å, respectively. Now, consider the case of a non-stoichiometric composition, the composition of which is shifted from the stoichiometric composition (Gd,Ce):(Al,Ga)=3:5. When Gd components exceed the stoichiometric composition, the Gd ions cannot be substituted into the (Al,Ga), sites since the ion radius of the Gd ion is large. Thus, these extra Gd ions are present in the garnet structure host crystal, the hetero phase of which is the GdAlO3 perovskite structure. On the other hand, when the Gd component is less than that of the stoichiometric composition, that is, when the (Al,Ga) components exceed the stoichiometric composition range, Al or Ga ions having a small radius can be substituted into the Gd sites and take on the garnet structure. From this consideration of the crystal structure, the property change of the light emission output and afterglow can be described as follows: When there is an excess of Gd component, a hetero phase (absorber) which does not emit light is generated in the host crystal along with an increase of the Gd components, and the light emission output is gently reduced. However, since the hetero phase GdAlO3 is not involved with light emission, it hardly at all affects the afterglow properties. On the other hand, when there is less of the Gd component than in the stoichiometric composition, the crystal structure retains the garnet structure. However, the Al or Gs ion deforms the structure of the crystal energy band and generates energy levels which allow electron transition to a forbidden band, since the Al or Gs ions which substitute for Gd sites are impurity ions in the garnet structure. It appears that, as a consequence, the afterglow becomes large, and the light emission output is also deteriorated, since the efficiency of energy transmission from the host crystal to the main luminescence center Ce3+ deteriorates. In this manner, in a phosphor including Ce as a luminescence element and a host crystal having a garnet structure including at least Gd, Al, Ga, and O, the luminescence properties and afterglow properties are greatly deteriorated even when the composition is slightly changed to have less (Gd+Ce) than in the stoichiometric composition (Gd+Ce)/(Al+Ga+Gd+Ce)=0.375. Theoretically, good scintillator properties should be obtained when the stoichiometric composition (Gd+Ce)/(Al+Ga+Gd+Ce)=0.375. However, in the process of manufacturing a scintillator plate, sintering easily causes the (Gd+Ce) components in this phosphor to decrease. Particularly, such influence becomes significant when a thick sintered body is employed. When a phosphor powder of the stoichiometric composition is employed, the phosphor properties of the sintered body are deteriorated. Therefore, in accordance with the present invention, the (Gd+Ce) component of the phosphor powder composition is made excessive by the amount with which good scintillator properties can be obtained, compensating for the composition misalignment effect by sintering. Thus, a phosphor element with low afterglow can be manufactured even when a thick sintered body is employed. The crystal form of the phosphor, according to this invention, is not particularly limited, and either a single crystal or a poly crystal can be employed. However, the poly crystal is preferable in terms of ease of manufacture and uniformity of properties. The poly crystal can be obtained by 1) the process of synthesizing the powder which is to be a raw material of the scintillator, and then 2) the process of sintering the powder. To obtain an oxide of the desired garnet structure, a smaller crystal-grain diameter of the synthesized powder is preferable, 10 μm or less being desirable. The methods of powder synthesis include 1) a method basically using a conventional oxide-mixing method, 2) a method using a liquid phase, such as a co-precipitation method and a sol-gel method, 3) a method based on an oxide-mixing method, in which the synthesized powder is finely divided by mechanical means, and the like. When using a conventional oxide-mixing method, the following manufacturing processes are feasible. A predetermined amount of powder, including such constituent metallic material as Gd2O3, Ce2O3, Ga2O3 in powder form, is weighed, and the powder is mixed in wet process using a ball mill, an automatic mortar, or the like. This mixed powder is sintered in 1000° C.˜1700° C. air or oxygen for several hours, thus resulting in the synthesized powder for a scintillator. If necessary, generation of a Gd—Ce—Al—Ga—O-system garnet structure can be improved by using, as a flux, a potassium compound, such as K2SO4, or a fluorine compound, such as BaF2. In the process using co-precipitation, synthesis can be performed as follows, for example: A predetermined amount of gadolinium nitrate, alumina nitrate, gallium nitrate, and cerium nitrate is weighed so as to make a compound nitrate water solution, and urea is made to coexist with nitrate ions, the urea concentration being fifteen times as much as the total metal-ion concentration. This water solution is heated to 70–100° C. to hydrolyze the urea, and a Gd—Al—Ga—O precursor is deposited. The deposition is repeatedly cleaned so as to lower the concentration of negative ions, which are useless, in the deposition to less than 1000 ppm, and it is dried at about 120° C. and subject to temporary sintering at about 1200° C., whereby the scintillator powder is made. The drying temperature should be 90° C. or more for good evaporation of moisture, and the temporary-sintering temperature should be 900° C. or more, temperatures at which the garnet structure is generated. Heat treatment in which the crystal grains of the generated garnet structure are developed must be avoided. The material used in the co-precipitation method is not limited to a nitrate, and a nitrate, sulfate, oxalate and the like of any kind of metal may be employed. Depending on the circumstances, several kinds of metallic salts may be mixed and used. Ammonium bicarbonate may also be used instead of urea. The Gd—Ce—Al—Ga—O precursor can also be deposited by adding to the compound metallic solution 1) urea and ammonium lactate, 2) ammonium bicarbonate and ammonia water, 3) ammonium sulfate and ammonia water, or 4) oxygenated water, as a masking agent, and ammonium sulfate and ammonia water. Finely dividing powder in mechanical manner is also a method suitable for obtaining the garnet structure by sintering. That is, as in the above-described oxide-mixing method, a predetermined amount of oxide of the metal which is to be a component of the composition is weighed and mixed by an automatic mortar for about thirty minutes. The mixed powder is subject to temporary sintering at around 1500° C., and it is mechanically grinded. A grinding method with high grinding energy, such as a ball mill, or preferably a planetary ball mill, should be employed. Thus, powder having a 0.01 to 0.5 μm grain diameter can be easily obtained. Sintering of the thus-synthesized powder can be carried out using a hot pressing method, an HIP method, a pressureless sintering method, simultaneous use of the pressureless sintering method and the HIP method, and the like. In any case, the relative density of the sintered body should be 99.0% or more, and preferably 99.5% or more. The “relative density” is a percentage of the actual density, the theoretical density of the material being 100. If the relative density becomes low, the scattering of light increases and the light transmission is extremely deteriorated, whereby sufficient light output cannot be obtained. Therefore, the relative density should be within the above-described range. When the hot pressing method is employed, the above-described synthesized powder is shaped with a mold with 500 kgf/cm2 pressure, set into a hot-press mold, and sintered under about 500 kgf/cm2 pressure in vacuum, air, or oxygen at a sintering temperature 1000° C.–1700° C. In this manner, a phosphor having 99.5% or more relative density can be easily obtained. In the HIP method, the synthesized powder is put into a metallic capsule made from iron, W, Mo, or the like, vacuum sealed, and sintered at about 1400° C. under 2000 atm pressure. In the pressureless sintering method, the synthesized powder is shaped with a mold under about 500 kgf/cm2 pressure, and is then subject to cold isostatic pressing (CIP) with about 3000 kgf/cm2 pressure. Then, it is sintered for several to several-ten hours at about 1400–1800° C. When the temperature is over 1800° C., the sample melts. When the temperature is under 1400° C., the sintering density becomes about 90% and sufficient sintering density cannot be obtained. To obtain a sintered body with 99.5% or more relative density by the pressureless sintering method, 1) powder to which a sintering auxiliary agent is added, or 2) minute powder of a sub-micron size, is preferably employed as the synthesized powder. Also, since pores are closed by pressureless sintering when the relative density of the sintered body is about 93.0% or more, a phosphor having 99.5% or more relative density can be easily obtained by employing the capsule-free HIP method in which a metallic capsule is not needed. In the phosphor obtained by the above-described method according to the present invention, the host crystal has a garnet structure, the material has its main luminescence spectrum peak at about 535 nm, and the damping factor of the afterglow 300 ms after stopping the excitation source is 1×10−4 or less. This oxide phosphor has a high light emission output and an extremely small afterglow, whereby it is suitable for use as a radiation detector of X-rays and the like, particularly for a radiation detector for a X-ray CT apparatus and a positron camera. As a light detector of the radiation detector according to the present invention, a PIN-type diode is employed. This photodiode has a high sensitivity, a fast response time, and a wavelength sensitivity through the visible ray area and the near infrared ray area, whereby the phosphor according to the present invention well matches the luminescence wavelength. Further, in the X-ray CT apparatus according to the present invention, the above-described radiation detector is employed as an X-ray detector. By employing such an X-ray detector, X-rays can be detected with high detecting efficiency, whereby the sensitivity thereof is greatly improved in comparison with an X-ray CT apparatus using a conventional scintillator (CdWO4, for example). Further, since the afterglow is extremely low, an image having a high image-quality and a high resolution can be obtained. An embodiment of an X-ray detector and an X-ray CT apparatus according to the present invention will be described with reference to the drawings. FIG. 1 shows one example of an X-ray detector 10 using scintillators according to the invention. In the X-ray detector 10, a scintillator 11 is adhered to a photodiode 13, and it is covered with a covering member 12 for preventing light emitted by the scintillators from leaking to the outside. X-rays pass through the covering member, which is made of aluminum or another material that reflects light. Incidentally, when applying the X-ray detector according to the invention to an X-ray CT apparatus, this X-ray detector is constructed such that several hundreds to several ten thousands of elements (detection channels) are arranged. FIG. 1 only shows three detection elements, which constitute one part of the detector. In this X-ray detector, the scintillator 11 includes a phosphor according to the invention, which has an extremely small afterglow and a high light emission output in comparison with a conventional scintillator. Since the scintillator 11 has its luminescence peak near the 535 nm wavelength, which is relatively close to the sensitivity wavelength of Si photodiodes, X-rays absorbed by the scintillator 11 are photoelectrically converted by the photodiodes with high efficiency. Therefore, the X-ray detector according to the invention exhibits an excellent performance, including a high sensitivity and an extremely small afterglow. FIG. 2 shows an outline of an X-ray CT apparatus according to the invention. This apparatus includes a gantry portion 18, an image reconstruction unit 22, and a monitor 23. The gantry portion 18 is provided with a rotating circular plate 19 having an opening 20 into which an object to be examined is inserted, an X-ray tube 16 mounted on this rotating circular plate, a collimator 17 which is mounted on the X-ray tube and controls the direction of X-ray irradiation, an X-ray detector 15 which is mounted on the rotating circular plate opposite to the X-ray tube, a detector circuit 21 for converting an X-ray dose detected by the X-ray detector 15 into a particular electric signal, and a scanning control circuit 24 for controlling rotation of the rotating circular plate and the width of X-ray irradiation. The X-ray detector 15 is constructed such that a number of detection elements, each of which combines the scintillator 11 and the photodiode 13 according to the invention, as shown in FIG. 1, are lined up along the circumference of the rotating circular plate 19. This X-ray detector 15 detects the dose of X-rays, which are radiated and pass through the object. The image reconstruction unit 22 includes an input device 25 for inputting the object's name, examination date, examination conditions, and the like, an image calculating circuit (not shown) for performing data-processing on measured data sent from the detector circuit 21, an image information addition unit (not shown) for adding information, such as the object's name, the examination date, and the examination conditions which were input by the input device 25, to CT images created by the image calculation circuit, and a display circuit (not shown) for adjusting the display gain of the CT images to which image information is added and output to the monitor 23. In the X-ray CT apparatus having such a structure, X-rays are irradiated from the X-ray tube 16 to the object, which is laid on a bed (not shown) provided in the opening 20. The X-rays are provided with directivity by the collimator 17, and they are detected by the X-ray detector 15. By rotating the rotating circular plate 19 around the object, X-rays are detected while changing the radiating direction of the X-rays, tomograms are created by the image reconstruction unit 22, and they are displayed on the monitor 23. Here, as the X-ray detector 15, a detector using the phosphor according to the invention, which has a low afterglow and a large emission intensity, is employed. Therefore, the image is not deteriorated due to afterglow, and images having a high image quality and a high resolution can be obtained. Hereinafter, various embodiments of the present invention will be described in detail. Gd2O3, Ce2(C2O4)3, Al2O3, and Ga2O3 were used as the material powder. The atomic ratios Ce/(Ce+Gd) and Ga/(Al+Ga) were fixed at 0.004 and 0.44, respectively. The material powder was weighed out so that the atomic ratio (Gd+Ce)/(Al+Ga+Gd+Ce) was 0.325 to 0.45, which corresponds to varying the composition along the thick line L in the composition diagram of FIG. 3. Next, the measured material powder, alumina balls, and ion exchange water were put into a vessel made of polyethylene, and they were mixed by a ball mill for twelve hours. This mixed powder was put into an evaporating dish and dried, and the dried powder was sized by a nylon sieve. An alumina crucible was filled with the sized powder, and it was sintered in 1500° C. oxygen for four hours. X-rays were irradiated from an X-ray source (120 kV, 0.5 mA) to the synthesized powder, and the afterglow and the luminescence intensity were measured. In this measurement, a detector using photodiodes is located 15 cm from the X-ray source, and the amount of light was measured. The afterglow is measured in terms of the damping factor 300 ms after stopping the X-rays, and the emission intensity is represented by a relative value. The amount of generated hetero-phase is calculated from powder X-ray diffraction tests. The hetero-phase amount is defined as the ratio between the strength of the main diffraction line of the garnet crystal structure, which is the matrix, and the strength of the main diffraction line of the perovskite crystal structure, which is the hetero-phase. FIG. 5 shows the result thereof. As is clear from the figure, the hetero-phase GdAlO3 is generated in proportion to the amount of composition misalignment from the stoichiometric composition when the atomic ratio (Gd+Ce)/(Al+Ga+Gd+Ce) is larger than 0.375. The afterglow properties are hardly changed at all, being stable at about 4×10−5, which is an extremely good value, as the atomic ratio is raised to 0.44. On the other hand, the luminescence properties are moderately deteriorated with an increase of the (Gd+Ce) component. When the atomic ratio (Gd+Ce)/(Al+Ga+Gd+Ce) exceeds 0.44, the light emission output becomes about 60% of the stoichiometric composition or less, and the ratio of the perovskite phase, which is the hetero-phase present in the phosphor, is over 50%, indicating that the synthesized powder having such an atomic ratio is not suitable as a phosphor material. On the other hand, when the (Gd+Ce) component is less than that in the stoichiometric composition, the afterglow properties and the luminescence properties are greatly deteriorated in response to a slight composition misalignment. Even when the (Gd+Ce)/(Al+Ga+Gd+Ce) composition is reduced by only 0.003 in comparison with the stoichiometric composition, the value of the afterglow becomes about a hundredfold, and the light emission output is decreased by about 30%. Although the (Gd+Ce)component is decreased, the crystal structure remains as a garnet structure, and only when (Gd+Ce)/(Al+Ga+Gd+Ce) is 0.325 is the hetero-phase finally generated. Gd2O3, Ce2(C2O4)3, Al2O3, and Ga2O3 were used as a material powder. The atomic ratios Ce/(Ce+Gd) and Ga/(Al+Ga) were fixed as 0.004 and 0.44, respectively. In this manner, a non-stoichiometric composition sample with (Gd+Ce)/(Al+Ga+Gd+Ce)=0.380 was manufactured. The material powder was measured out according to the composition in Chart 1. Next, the measured material powder, alumina balls, and ion exchange water were put into a vessel made of polyethylene, and they were mixed by a ball mill for about 12 hours. The mixed powder was put into an evaporating dish and dried. The dried powder was sized by a nylon sieve. An alumina crucible was filled with the sized powder, and it was sintered in 1500° C. oxygen for about four hours. This synthesized powder was molded using a mold having 160 mm inner diameter with 500 kgf/cm2 pressure, and a molded body was thus made. It was then set in a hot press die and hot-press sintered in a vacuum at 1475° C. for four hours under 500 kgf/cm2 pressure, and sintered bodies having a 4.7 mm (Embodiment 2) thickness and having a 15 mm (Embodiment 3) thickness were thus obtained. The relative density of these sintered bodies were both 99.9% or more. The sintered bodies that were obtained by the above-described method were sliced in the plate-thickness direction, one (Embodiment 2) or five (Embodiment 3) wafers having a 160 mm inner diameter were cut out, and the wafers were cut into a predetermined length and machined so as to have a 108 mm width. Thus, scintillator plates were manufactured in this way. These scintillator plates were annealed in oxygen for four hours, and the afterglow and the emission intensity were then measured. The result thereof is set forth in the chart 1. Incidentally, the measurement of the afterglow and the emission intensity is performed for detectors made by combining the above-manufactured scintillator plates and photodiodes, and the detector is located 110 cm away from the X-ray source (120 kV, 150 mA). The afterglow is measured in terms of the damping factor 300 ms after stopping the X-rays, and the emission intensity is represented by a relative value with the emission intensity of CdWO4 set as 1. Sintered bodies respectively having a 4.7 mm width, a 10.0 mm width, and a 15.0 mm width were manufactured in the same manner as those of Embodiments 2 and 3, except that the atomic ratio (Gd+Ce)/(Al+Ga+Gd+Ce) of the starting material powder is 0.375. One, three and five wafers having a 160 mm inner diameter were cut out, respectively, from the above-obtained sintered bodies, and scintillator plates (Comparative Examples 1 to 3) were thus produced. These scintillator plates were also annealed in the same manner as Embodiments 2 and 3, and the afterglow and the emission intensity were measured. The result thereof is set forth in the chart 1. Incidentally, in the interior of the respective scintillators, a great difference in the afterglow and emission intensity was not found between them in the plate-thickness direction and in the diameter direction. CHART 1SinteredEmissionBody(Gd + Ce)/IntensityThickness(Gd + Ce +Afterglow(versus(mm)Al + Ga)GdCeAlGa(300 ms)CsWO4)Embodiment 24.70.3803.030.012.782.181.2 × 10−52.3Embodiment 315.00.3803.030.012.782.181.4 × 10−52.3Comp. Ex. 14.70.3752.990.012.802.202.0 × 10−52.2Comp. Ex. 210.00.3752.990.012.802.203.0 × 10−42.0Comp. Ex. 315.00.3752.990.012.802.209.50 × 10−4 1.7 In the samples of the stoichiometric composition (Comparative Examples 1 to 3), the afterglow was drastically increased as the plate thickness was increased. When the plate thickness was 15 mm, the value of the afterglow was increased by double digits. The emission intensity also deteriorated. On the other hand, in the samples of the non-stoichiometric composition (Embodiment 2 and 3), the afterglow was not increased and the emission intensity also was not deteriorated, even when the plate thickness was increased from 4.7 mm to 15 mm. Incidentally, the difference between 0.375 and 0.38 in the atomic ratios of (Gd+Ce)/(Al+Ga+Gd+Ce), being the measured out compositions, is not the result of a mere measuring error in manufacturing the material or an error occurring when the material powder absorbs moisture. This is because such errors are kept within 0.001 at the maximum. Sintered bodies were manufactured in the same manner as those in Embodiments 2 and 3, except that the atomic ratio of Gd, Ce, Al, and Ga was as shown in Chart 2. Though the atomic ratio Ga/(Al+Ga) is 0.44 in the Comparative Examples 1 to 3 and Embodiments 2 and 3, that in the Comparative Examples 4 to 7 and in Embodiments 4 to 8 is 0.70, and that in the Comparative Examples 8 to 11 and in Embodiments 9 to 13 is 0.30. The Ce concentration is fixed so that the atomic ratio Ce/(Ce+Gd) is 0.004, and the ratio (Gd+Ce)/(Al+Ga+Gd+Ce) is varied. The thickness of all sintered bodies is 15 mm. The obtained sintered bodies were sliced in the plate-thickness direction. After cutting, the sintered bodies were machined into plates with a thickness of 1.8 mm, and they were annealed at 1300° C. The hetero-phase amount in these scintillator plates was measured as the intensity ratio between the main peak of the diffraction line of the garnet structure and that of the hetero phase, which can be obtained by use of an X-ray diffraction apparatus. Detectors were formed by combining the obtained scintillator plates and photodiodes, they were located 110 cm from an X-ray source (120 kV, 150 mA), as in Embodiments 2 and 3, and the emission intensity and the afterglow were thus measured. The emission intensity was represented by a relative value, with the value of CdWO4 set to be 1, and the afterglow was measured in terms of the damping factor 300 ms after stopping the X-rays. The result thereof is set forth in Chart 2 and Chart 3. CHART 2Hetero-Emission(Gd + Ce)/phaseIntensity(Gd + Ce +amountAfterglow(versusAl + Ga)GdCeAlGa(%)(300 ms)CdWO4)Comp. Ex. 40.352.790.011.563.6401.5 × 10−31.3Comp. Ex. 50.372.950.011.513.5301.5 × 10−31.3Comp. Ex. 60.3752.990.011.503.5001.9 × 10−41.8Embodiment 40.3763.000.011.503.4901.0 × 10−51.9Embodiment 50.3783.010.011.493.4801.0 × 10−51.9Embodiment 60.393.110.011.463.4291.0 × 10−51.8Embodiment 70.413.270.011.423.30241.5 × 10−51.6Embodiment 80.443.510.011.343.14392.5 × 10−51.5Comp. Ex. 70.463.670.011.303.02583.5 × 10−51.2 CHART 3Emission(Gd + Ce)/Hetero-Intensity(Gd + Ce +PhaseAfterglow(versusAl + Ga)GdCeAlGaAmount(300 ms)CdWO4)Comp. Ex. 80.352.790.013.641.5601.5 × 10−31.1Comp. Ex. 90.372.950.013.531.5101.5 × 10−31.2Comp. Ex. 100.3752.990.013.501.5009.8 × 10−51.6Embodiment 90.3763.000.013.491.5001.0 × 10−51.7Embodiment 100.3783.010.013.481.4901.0 × 10−51.7Embodiment 110.393.110.013.421.4681.0 × 10−51.6Embodiment 120.413.270.013.301.42292.0 × 10−51.4Embodiment 130.443.510.013.141.34403.5 × 10−51.3Comp. Ex. 110.463.670.013.021.30583.8 × 10−51.1 Even when the atomic ratio Ga/(Al+Ga) is 0.70 or 0.30, the hetero-phase amount is small and the emission intensity is relatively large as long as the atomic ratio (Gd+Ce)/(Al+Ga+Gd+Ce) is 0.375 or less (Comparative Examples 4 to 6, 8 to 10). However, since the thus obtained material had a damping factor of 10−3 to 10−4 of the afterglow after 300 ms, it was shown that the sintered body used therein became large and was not suitable as a scintillator material. When the atomic ratio (Gd+Ce)/(Al+Ga+Gd+Ce) was over 0.44 (Comparative Example 7 and 11), the hetero-phase amount was over 50%, and the emission intensity also was lowered. On the other hand, the scintillators according to Embodiments 4 to 13 of the present invention have an extremely short afterglow and a high emission intensity, whereby it is understood that they have excellent scintillator properties. Sintered bodies (15 mm thickness) of Embodiments 14 to 17 and of the Comparative Examples 12 and 13 were manufactured in the same manner as those in Embodiments 2 and 3, except that the atomic ratio (Ge+Ce)/(Al+Ga+Gd+Ce) was set to be 0.38, the atomic ratio Ce/(Ce+Gd) as the Ce concentration was set to be 0.004, and the atomic ratio Ga/(Al+Ga) was varied from 0.0 to 1.0. These sintered bodies were sliced in the plate-thickness direction. After cutting, they were machined so as to make their thickness 1.8 mm. Scintillator plates were thus made, and they were annealed at 1300° C. Detectors were formed by combining the obtained scintillator plates and photodiodes, they were located 110 cm from an X-ray source (120 kV, 150 mA), and the afterglow and the emission intensity were thus measured. The afterglow was measured in terms of the damping factor 300 ms after stopping X-rays, and the emission intensity is represented by a relative value with respect to the value 1 of CdWO4. The results thereof are set forth in Chart 4. CHART 4Emission(Gd + Ce)/Intensity(Gd + Ce +Ga/Afterglow(versusAl + Ga)(Al + Ga)GdCeAlGa(300 ms)CdWO4)Comp. Ex. 120.380.003.030.014.960.00—0.0Embodiment 140.380.053.030.014.710.252.0 × 10−51.2Embodiment 150.380.303.030.013.471.491.5 × 10−52.1Embodiment 160.380.603.030.011.982.982.0 × 10−52.0Embodiment 170.380.903.030.010.504.461.5 × 10−51.3Comp. Ex. 130.381.003.030.010.004.96—0.0 From the chart, it is seen that, although the scintillators of Embodiments 14 to 17 have excellent afterglow and emission intensity properties, the emission intensity of the scintillators whose atomic ratio Ga/(Al+Ga) is 0 or 1.0 (Comparative Example 12 and Comparative Example 13), is much lower, and the sintered bodies thereof are not suitable for use as a scintillator. The scintillators of Embodiments 14 and 15 were manufactured in the same manner as those of Embodiments 2 and 3, except that the atomic ratio Ce/(Ce+Gd) is varied from 0.002 to 0.04. Incidentally, the atomic ratios (Gd+Ce)/(Al+Ga+Gd+Ce) and Ga/(Al+Ga) are fixed, respectively, to 0.38 and 0.44. The thickness of the sintered bodies is 15 mm. The result thereof is shown in Chart 5. CHART 5Emission(Gd + Ce)/Intensity(Gd + Ce +Ce/Afterglow(versusAl + Ga)(Ce + Gd)GdCeAlGa(300 ms)(CdWO4)Comp. Ex. 140.380.00023.040.00062.782.183.5 × 10−40.8Embodiment 180.380.00053.040.00152.782.184.8 × 10−51.6Embodiment 190.380.0013.040.00302.782.181.5 × 10−52.1Embodiment 200.380.0053.020.01522.782.181.0 × 10−52.0Embodiment 210.380.022.980.06082.782.183.0 × 10−51.5Comp. Ex. 150.380.042.920.12162.782.182.5 × 10−50.5 From the chart, it is seen that the scintillators of Embodiments 18 to 21 have excellent properties of afterglow and emission intensity, and the scintillators of the Comparative Examples 14 and 15, where the atomic ratio Ce/(Ce+Gd) is respectively less than 0.0005 and more than 0.02, are not suitable as the required scintillator because the emission intensities thereof are extremely low. According to the present invention, a phosphor having an extremely low afterglow and a high luminescence efficiency can be provided by reducing the composition misalignment occurring during sintering, which is a drawback of a phosphor of (Gd1-xCex)3Al5-yGayO12 composition. Further, by using this phosphor as a scintillator for a radiation detector having a light detector, a radiation detector having a low afterglow and a high output can be obtained. By applying this radiation detector to an X-ray CT apparatus, a tomogram having a high resolution and a high quality can be obtained. |
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059828382 | abstract | The present invention relates to a method and portable apparatus which is used to detect substances, such as explosives and drugs, by neutron irradiation. The apparatus has a portable neutron generating probe and corresponding controllers and data collection computers. The probe emits neutrons in order to interrogate an object. The probe also contains gamma ray detectors for the collection of gamma rays from fast neutron, thermal neutron and neutron activation reactions. Data collected from these detectors is sent to the computer for data de-convolution then object identification in order to determine whether the object being interrogated contains explosives or illicit contraband. |
claims | 1. A drive device for an in-core neutron instrumentation system, the drive device driving a neutron detector of the in-core neutron instrumentation system for measuring neutrons in a nuclear reactor, comprising:a winding drum for winding a cable having the neutron detector mounted at a tip end portion thereof;a drive motor for driving the winding drum for winding/extending the cable;a base member on which the winding drum and the drive motor are mounted;a cover member mounted on the base member and housing the winding drum and the drive motor therein; anda safety circuit provided between a power supply of the drive motor and the drive motor, and for switching between a conductive state in which power can be supplied to the drive motor and a cut-off state in which power cannot be supplied thereto depending on a mounted state of the cover member,wherein the cover member comprises a plurality of cover wall surface portions surrounding a periphery of the base member, and is equipped with a cover locking mechanism portion for defining an order of mounting the cover wall surface portions on the base member, and whereinthe safety circuit is provided on one of the plurality of cover wall surface portions, andthe cover locking mechanism portion is provided so that the one cover wall surface portion on which the safety circuit is provided is finally mounted in the order of mounting the plurality of cover wall surface portions on the base member. 2. The drive device for the in-core neutron instrumentation system according to claim 1, wherein the safety circuit is controlled to be in the conductive state in response to a state in which the cover member is mounted on the base member, and is controlled to be in the cut-off state in response to a state in which the cover member is not mounted on the base member. 3. The drive device for the in-core neutron instrumentation system according to claim 1, whereinthe cover member comprises a cover upper surface portion for covering an upper portion of a space surrounded by the plurality of cover wall surface portions,the cover wall surface portions and the cover upper surface portion are held by a frame which is fixed to the base member and provided in abutting regions between adjacent two of the cover wall surface portions and in abutting regions between the cover wall surface portions and the cover upper surface portion, andthe cover locking mechanism portion is mounted on the frame. 4. The drive device for the in-core neutron instrumentation system according to claim 1, whereinthe cover member comprises a cover upper surface portion for covering an upper portion of a space surrounded by the plurality of cover wall surface portions, andthe safety circuit is configured to include a series circuit which is brought into the conductive state in response to the whole cover member being mounted on the base member. 5. A drive device for an in-core neutron instrumentation system, the drive device driving a neutron detector of the in-core neutron instrumentation system for measuring neutrons in a nuclear reactor, comprising:a winding drum for winding a cable having the neutron detector mounted at a tip end portion thereof;a drive motor for driving the winding drum for winding/extending the cable;a base member on which the winding drum and the drive motor are mounted;a cover member mounted on the base member and housing the winding drum and the drive motor therein; anda safety circuit provided between a power supply of the drive motor and the drive motor, and for switching between a conductive state in which power can be supplied to the drive motor and a cut-off state in which power cannot be supplied thereto depending on a mounted state of the cover member,wherein the cover member comprises a plurality of cover wall surface portions surrounding a periphery of the base member, and is equipped with a cover locking mechanism portion for defining an order of mounting the cover wall surface portions on the base member, and whereinthe safety circuit is provided on one of the plurality of cover wall surface portions, andthe cover locking mechanism portion is provided so that the one cover wall surface portion on which the safety circuit is provided is first removed in an order of removing the plurality of cover wall surface portions from the base member. 6. The drive device for the in-core neutron instrumentation system according to claim 5, wherein the safety circuit is controlled to be in the conductive state in response to a state in which the cover member is mounted on the base member, and is controlled to be in the cut-off state in response to a state in which the cover member is not mounted on the base member. 7. The drive device for the in-core neutron instrumentation system according to claim 5, whereinthe cover member comprises a cover upper surface portion for covering an upper portion of a space surrounded by the plurality of cover wall surface portions,the cover wall surface portions and the cover upper surface portion are held by a frame which is fixed to the base member and provided in abutting regions between adjacent two of the cover wall surface portions and in abutting regions between the cover wall surface portions and the cover upper surface portion, andthe cover locking mechanism portion is mounted on the frame. 8. The drive device for the in-core neutron instrumentation system according to claim 5, whereinthe cover member comprises a cover upper surface portion for covering an upper portion of a space surrounded by the plurality of cover wall surface portions, andthe safety circuit is configured to include a series circuit which is brought into the conductive state in response to the whole cover member being mounted on the base member. |
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055984533 | abstract | An X-ray apparatus for collecting X-ray transmission data of a subject from a plurality of directions to generate an X-ray transmission image or X-ray CT image of the subject, which includes an X-ray generator for generating an X-ray, an X-ray detector for detecting a transmission X-ray after the X-ray generated by the X-ray generator is transmitted through the subject, a rotation unit for rotating an imaging unit including the X-ray generator and the X-ray detector around the subject, a data collector for converting an output signal of the X-ray detector to a digital signal and collecting the digital signal, a signal processor for subjecting data collected by the data collector to a signal processing operation, a display for displaying thereon as an image the data collected by the data collector and the data subjected by the signal processor to the signal processing operation, and a position change unit for moving a relative position of a rotation center of the imaging unit and the subject in a direction parallel to a rotation plane of the rotation, and wherein the imaging unit is rotated by the rotator around the subject and at the same time the relative position is changed by the position change unit in a direction parallel to the rotation plane to perform X-ray fluoroscopic or radiographic operation or CT scan. |
description | This invention pertains generally to methods and devices for the insertion and removal of radioactive isotopes into and out of a nuclear core and, more particularly, to the insertion and removal of such isotopes into and out of a commercial nuclear reactor on a mass production basis without reducing the reactor's facilities ability to generate electricity. The commercial production of radioactive isotopes for medical and other commercial enterprises, such as Radioisotope Thermal Generators (RTG), is a process which is limited by the very high costs associated with developing the neutron source infrastructure required to create commercial quantities of the useful isotopes. This makes the useful applications of these radioactive isotopes very expensive and subject to extreme supply and cost fluctuations due to actual or perceived potential interruptions at the very limited number of isotope production facilities available. The human cost associated with this situation is that most people are not able to afford the cost of the medical benefits that can be provided by the large number of available radioactive isotope diagnostic and treatment modalities. Furthermore, the reactors that are currently used to produce the radioisotopes that are processed to produce radio-pharmaceuticals are very old, and continued operation requires very expensive upgrades that appear to provide poor return on investment. Consequently, the reactor resources required to maintain existing production capability is disappearing. The fundamental issue to be addressed is the loss of medical radioisotope production capability due to obsolescence issues in the existing medical radioisotope production infrastructure that will lead to a shortage of the radioisotopes needed to diagnose and treat serious medical issues. Accordingly, a need exists for an alternative, and preferably less expensive, way of producing radioisotopes. A number of operating nuclear reactors used in commercial electrical generation facilities employ a moveable in-core detector system such as the one described in U.S. Pat. No. 3,932,211, to periodically measure the axial and radial power distribution within the core. The moveable detector system generally comprises four, five or six detector/drive assemblies, depending upon the size of the plant (two, three or four loops), which are interconnected in such a fashion that they can assess various combinations of in-core flux thimbles. To obtain the thimble interconnection capability, each detector has associated with it a five or six-path and ten or fifteen-path rotary mechanical transfer device. A core map is made by selecting, by way of the transfer devices, particular thimbles through which the detectors are driven. To minimize mapping time, each detector is capable of being run at high speed (72 feet per minute) from its withdrawn position to a point just below the core. At this point, the detector speed is reduced to 12 feet per minute and the detector traversed to the top of the core, direction reversed, and the detector traversed to the bottom of the core. The detector speed is then increased to 72 feet per minute and the detector is moved to its withdrawn position. A new flux thimble is selected for mapping by rotating the transfer devices and the above procedure repeated. FIG. 1 shows the basic system for the insertion of the movable miniature detectors. Retractable thimbles 10, into which the miniature detectors 12 are driven, take the routes approximately as shown. The thimbles are inserted into the reactor core 14 through conduits extending from the bottom of the reactor vessel 16 through the concrete shield area 18 and then up to a thimble seal table 20. Since the movable detector thimbles are closed at the leading (reactor) end, they are dry inside. The thimbles, thus, serve as a pressure barrier between the reactor water pressure (2500 psig design) and the atmosphere. Mechanical seals between the retractable thimbles and the conduits are provided at the seal table 20. The conduits 22 are essentially extensions of the reactor vessel 16, with the thimbles allowing the insertion of the in-core instrumentation movable miniature detectors. During operation, the thimbles 10 are stationary and will be retracted only under depressurized conditions during refueling or maintenance operations. Withdrawal of a thimble to the bottom of the reactor vessel is also possible if work is required on the vessel internals. The drive system for insertion of the miniature detectors includes, basically, drive units 24, limit switch assemblies 26, five-path rotary transfer devices 28, 10-path rotary transfer devices 30, and isolation valves 32, as shown. Each drive unit pushes a hollow helical wrap drive cable into the core with a miniature detector attached to the leading end of the cable and a small diameter coaxial cable, which communicates the detector output, threaded through the hollow center back to the trailing end of the drive cable. The use of the moveable in-core detector system flux thimbles 10 for the production of irradiation desired neutron activation and transmutation products, such as isotopes used in medical procedures, requires a means to insert and withdraw the material to be irradiated from inside the flux thimbles located in the reactor core 14. Preferably, the means used minimizes the potential for radiation exposure to personnel during the production process and also minimizes the amount of radioactive waste generated during this process. In order to precisely monitor the neutron exposure received by the target material to ensure the amount of activation or transmutation product being produced is adequate, it is necessary for the device to allow an indication of neutron flux in the vicinity of the target material to be continuously measured. Ideally, the means used would be compatible with systems currently used to insert and withdraw sensors within the core of commercial nuclear reactors. Co-pending U.S. patent application Ser. No. 15/210,231, entitled Irradiation Target Handling Device, filed Jul. 14, 2016, describes an Isotope Production Cable Assembly that satisfies all the important considerations described above for the production of medical isotopes that need core exposure for less than a full fuel cycle. There are other commercially valuable radioisotopes that are produced via neutron transmutation that require multiple neuron induced transmutation reactions to occur in order to produce the desired radioisotope product, or are derived from materials having a very low neutron interaction cross section, such as Co-60, W-188, Ni-63, Bi-213 and Ac-225. These isotopes require a core residence time of a fuel cycle or more. Commercial power reactors have an abundance of neutrons that do not significantly contribute to the heat output from the reactor used to generate electrical power. This invention describes a process and associated hardware that may be used to utilize the neutron environment in a commercial nuclear reactor to produce commercially valuable quantities of radioisotopes that require long-term neutron exposure, i.e., a fuel cycle or longer, or short term exposure, i.e., less than one fuel cycle, with minimal impact on reactor operations and operating costs. The hardware and methodology described in U.S. patent application Ser. No. 15/341,478, filed Nov. 2, 2016, will enable the production of radioisotopes that require relatively long residence times in the core, currently produced in outdated isotope production reactors, using the foregoing moveable in-core detector system equipment without interfering with the functionality of the moveable in-core detector system power distribution measurement process. There is still a further need for a more efficient radioisotope production process that can produce radioisotopes in commercial nuclear reactors on a mass production scale, without negatively impacting the electrical power output of those commercial facilities. It is an object of this invention to satisfy that need. This and other objects are achieved, in accordance with this invention, with an irradiation target handling system having an isotope production cable assembly comprising a target holder drive cable constructed to be compatible with conduits of an existing nuclear reactor moveable in-core detector system that convey in-core detectors from a detector drive system to and through instrument thimbles within a reactor core. The target holder drive cable has a remotely controlled one of a male or female coupling on a leading end of the drive cable. A target holder drive cable drive motor unit is provided separate from and independent of the detector drive unit on the existing nuclear reactor moveable in-core detector system. The target holder drive cable drive motor unit is configured to drive the target holder drive cable into and out of the core and is structured to drive the target holder drive cable into and through the conduits, a first multipath selector and a second multipath selector on the existing nuclear reactor moveable in-core detector system. A specimen target holder is provided having another of the male or female coupling on a trailing end of the specimen target holder with the another of the male or female coupling configured to mate with the one of the male or female coupling on the leading end of the target holder drive cable. A third multipath selector is connected to and structured to receive an input from an outlet path on the second multipath selector and provides a first output to a new specimen attachment location, a second output to an irradiated specimen offloading location and a third output to the core. In one embodiment the specimen target holder has a radial projection extending from or through an outside wall of the specimen target holder into contact with an interior wall of an instrument thimble in the reactor core, into which the specimen target holder is driven by the target holder drive cable, which maintains an axial position of the specimen holder within the instrument thimble, when the specimen holder is detached from the drive cable. Preferably, the one of the male or female coupling is configured to move the radial projection away from the interior wall of the instrument thimble when coupled to another of the male or female coupling on the specimen target holder. In still another embodiment the irradiation target handling system includes an axial positioning device attached to the specimen target holder for determining when the specimen target holder achieves a preselected axial position within an instrument thimble within the core, which the specimen target holder is driven into by the drive cable. Preferably, the instrument thimbles have a closed upper end and a leading end of the specimen target holder has an axial projection that is sized to contact an interior of the closed upper end of the instrument thimble into which the specimen target holder is driven. In one such embodiment the length of the axial projection is a wire having an adjustable length. Desirably, the target holder drive cable enters the conduits through a “Y” connection with one leg of the “Y” connected to the target holder drive cable drive motor unit and a second leg of the “Y” connected to the detector drive unit. The invention also contemplates a method of irradiating multiple specimens within a core of a nuclear reactor that has a moveable in-core, radiation detector flux mapping system, wherein the core comprises a plurality of fuel assemblies respectively having instrument thimbles into which a radiation detector of the flux mapping system can be inserted and travel through. The method comprises the step of inserting a first specimen holder containing a first specimen at a lead end of a first drive cable driven by a first drive unit, into a first instrument thimble in the core. Next, the method remotely detaches the first drive cable from the first specimen holder and fixes an axial position of the first specimen holder within the first instrument thimble. Then the first drive cable is withdrawn from the reactor. Next, a second specimen holder containing a second specimen is attached to the lead end of the first drive cable driven by the first drive unit. The second specimen holder containing the second specimen is then inserted into a second instrument thimble in the core. The first drive cable is next remotely detached from the second specimen holder and the second specimen holder is fixed at an axial position that it was driven to within the second instrument thimble. The next step withdraws the first drive cable from the reactor. In between the withdrawing step and the second inserting step, the method inserts a moveable in-core radiation detector from the moveable in-core detector radiation flux mapping system, attached to a second drive cable driven by a second drive unit, into and through a third instrument thimble and withdraws the moveable in-core radiation detector from the reactor after performing a flux mapping exercise. In one embodiment of the method, the inserting steps insert specimen holders into as many as half the instrument thimbles accessible by the flux mapping system for simultaneous irradiation at a time when a flux map is to be conducted. Preferably, the steps of fixing the axial position of the specimen holders within the respective instrument thimbles includes the steps of determining when the respective specimen holders are at a preselected axial position within the corresponding instrument thimbles. In one such embodiment, the step of withdrawing the first drive cable from the reactor comprises withdrawing the first drive cable out of the moveable in-core, radiation detector flux mapping system prior to the running of a flux map. To accomplish the foregoing objectives, this invention modifies the traditional flux mapping system described above with respect to FIG. 1, as shown in FIG. 2. FIG. 2 shows a portion of the moveable in-core detector flux mapping system containing the detector drive unit 24, the five-path transfer device 28, the ten-path transfer device 30 and the seal table 20 in schematic form with the incidental components, like the limit switches, safety switches and isolation valves omitted. Also shown in FIG. 2 are the core components of the modifications introduced by this invention to the moveable in-core detector flux mapping system, to convert the moveable in-core detector flux mapping system into a radioisotope mass production facility, without compromising the flux mapping function. In accordance with this invention a specimen holder cable drive unit 34 is provided that is distinct and independent of the detector drive unit 24. The specimen holder cable drive unit 34 drives a specimen holder drive cable 36 that has a specimen holder 48 detachably attached to the lead end of the specimen holder drive cable 36. The specimen holder 48 is shown in and will be described in more detail with regard to FIG. 3. It should also be appreciated that the specimen holder cable drive unit 34 and the specimen holder cable 36 may be configured the same as the detector motor drive unit 24 and the detector drive cable 50, though other configurations are also compatible with this invention. The specimen holder drive cable 36 is fed into the conduits of the moveable detector in-core flux mapping system through a “Y” connection 38 that communicates with the input to the five-path transfer device 28. One of the outputs of the five-path transfer device similarly feeds the input to the ten-path transfer device 30, one of the outputs 52 of which feeds a new three-path transfer device 40. One output of the three-path transfer device feeds a new specimen attachment point 42, at which a new specimen holder and specimen can be attached to the specimen holder drive cable; a second output of the three-path transfer device feeds a specimen holder catcher 44 in which the specimen holder can be offloaded; and a third output of the three-path transfer device provides a path to the core 54. It should be appreciated that while five-path, ten-path and three-path transfer devices are disclosed these devices may have as many paths as necessary to access the desired locations within the core and currently five-path and six-path devices 28 and ten-path and fifteen-path devices 30 are in use or planned for use, depending on the size of the core. FIG. 3 shows the lead end of the specimen holder drive cable 36 and the specimen holder 48. The specimen holder drive cable 36 has a spiral wire wrap 56 that mates with drive gears in the specimen holder drive motor unit 34 to advance and withdraw the specimen holder drive cable 36 through the conduits of the flux mapping system. At the lead end of the specimen holder drive cable 36 is a remotely operated male coupling component 58 that fits within a female coupling component 60 on the specimen holder 48. The male coupling component 58 has a remotely operated pneumatic latch plug 62 that when fully activated in its extended position fits within an annular groove 64 in the female coupling component 60. The latch plug 62 is shown in more detail in FIG. 3A, in the activated position, and includes an unlatching spring 66 that retracts the latch plug 62 when the pneumatic pressure supplied through the pneumatic fluid supply channel 70 is released. The pneumatic fluid is supplied from a pneumatic fluid supply reservoir 46, shown in FIG. 2, through the pneumatic fluid supply channel 70 that runs through the center of the specimen holder drive cable 36. A retaining clip 68 prevents the latch plug 62 from leaving the channel in which it travels. The specimen holder 48 has a payload chamber 72 that houses the specimen to be irradiated and two or more positioning tabs 76 that extend from an interior of the specimen holder housing 74, through the specimen holder housing and up against an interior surface of a fuel assembly instrument thimble in which the specimen holder 48 is to be inserted, to hold the specimen holder in position, by friction, when it is remotely disconnected from the specimen holder drive cable 36. The positioning tabs 76 are biased in a fully extended position and are rotated out of contact with the side walls of the instrument thimble by the male coupling component 58 when the male coupling component is fully inserted into the female coupling component 60. An end view of the positioning tabs 76 is shown in FIG. 3B. The specimen holder 48 also has an adjustable positioning cable 78 which extends out the lead end of the specimen holder 48. The desired axial position of the specimen within the instrument thimble is determined in advance of inserting the specimen into the moveable in-core detector flux mapping system and the length of the positioning cable 78 is adjusted so its lead end abuts the closed upper end of the instrument thimble when the specimen is at the desired position. Thus, in between flux map runs, which are typically conducted once a quarter, the moveable in-core detector flux mapping system is available to insert isotopes into and harvest isotopes from all of the instrument thimbles in a reactor core accessible to the flux mapping system, so long as at least fifty percent of those thimbles are unoccupied at the time a flux map is to be run. Prior to a flux mapping run, the specimen holder drive cable 36 has to be withdrawn above the “Y’ connection 38 to provide the miniature detector access to the five-path transfer device 28. Similarly, once a flux mapping run is completed, the miniature detector needs to be withdrawn above the “Y” connection to provide the specimen holder drive cable 36 access to the five-path transfer device 28. It should be appreciated that a typical reactor facility employing a moveable in-core flux mapping system has four, five or six parallel, interconnected trains of detectors whose detector drive cables can be run simultaneously so long as they are routed through different conduits to the core. In accordance with this invention each one of the detector trains can be provided with its own specimen holder cable drive unit that are individually programmed to plant isotopes at different desired locations within the core. Thus, this invention provides modifications to an existing moveable in-core detector system and a method to perform the following functions that: (i) enables the insertion of specially configured specimens through a specially configured access to the existing multi-path transfer devices from one or more detector drive trains that enable the specimen to be inserted into a desired radial reactor core location that can be reached through the existing multi-path routing options; (ii) enables the specimen to be inserted into the desired available core location at a predetermined axial position inside the moveable in-core detector system instrument thimble relative to the top of the active fuel in the desired fuel assembly; (iii) enables the specimen holder drive cable to be disconnected from the specimen holder and withdrawn from the reactor above the multipath transfer devices with the axial position of the specimen in the reactor fixed by mechanical features on the specimen holder side of the specimen holder drive cable connector; (iv) enables the specimen holder drive cable to be inserted through a specific existing multi-path transfer device position selection to another specially configured transfer device, located downstream of the existing multi-path transfer devices (hereafter referred to as the lower path selector), that has a position that enables the specimen holder drive cable end to reach a location that enables the specimen holder drive cable to have another specimen holder payload attached; (v) enables a new specimen to be withdrawn above the existing multi-path transfer devices and then repeat the above steps 1 through 4 until all the desired specimens are “planted” in the reactor core as planned; (vi) enables the specimen holder drive cable to be inserted into a planted specimen location so that the mating portions of the specimen holder drive cable connector are brought together to enable the latching plugs on the drive cable side of the connector to be activated using a pneumatic fluid, such as nitrogen, to pressurize the pneumatic fluid supply channel so the latch plugs insert into the latch channel located on the specimen side of the connector so that the specimen holder can be withdrawn, or “harvested,” following completion of the desired irradiation levels; (vii) enables the harvested specimen to be withdrawn through the ten-path selector device where the specimen holder latch to the drive cable is released by reducing the applied pneumatic fluid pressure, and then it is inserted through the lower path selector position that enables insertion of the specimen holder until it is captured by a device designed to coil the specimen holder with the specimen payload, to fit within the payload bay of a radioactive material transfer cask used for transportation of the specimen to a processing facility; (viii) enables the specimen holder cable to be positioned as described in step 4, above, and repeat steps 1 through 5 as desired; and (ix) enables steps 1 through 8 to be repeated as desired. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular embodiments disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the appended claims and any and all equivalents thereof. |
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abstract | Systems, devices, and methods are described for providing, among other things, an intra-oral x-ray imaging system configured to reduce patient exposure to x-rays, reduce amount of scatter, transmission, or re-radiation during imaging, or improve x-ray image quality. In an embodiment, an intra-oral x-ray imaging system includes an intra-oral x-ray sensor configured to communicate intra-oral x-ray sensor position information or intra-oral x-ray sensor orientation information to a remote x-ray source. |
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040381331 | summary | Prior art, gas-cooled fast breeder reactors commonly include a concrete pressure vessel containing a reactor core formed by a plurality of laterally adjacent fuel assemblies, each assembly comprising a vertically elongated casing containing a bundle of fuel rods, the casing having an upper end and extending downwardly therefrom and having an open lower end extending below the fuel rods. Each casing has a vertical tubular suspension rod having an upper end with a supported connection with the concrete pressure vessel top and a lower end with a connection to the casing's upper end so that the casing is suspended thereby. At least one of the connections of the tubular suspension rod is made releasable so that any one of the fuel assemblies can be lowered from the core for servicing of the core. These vertical tubular suspension rods form the primary suspensions for the fuel assemblies. If any one of these suspensions fail so as to accidentally release an assembly during operation of the reactor, the assembly can drop from the core. These primary suspensions are designed to provide a redundancy of safety, but a failure by one or more of the suspensions is at least hypothetically possible. During operation of the reactor the flow of gas coolant is downwardly through the fuel element casings and it is desirable to determine the operating temperatures of each of the fuel assemblies individually. Therefore, each fuel element has an instrumentation tube having an upper end releasably fixed above the casing, this upper end normally projecting upwardly through the top of the concrete pressure vessel within an external pressure tube providing for the passage of electrical lines in a pressure-tight manner to instrumentation external of the reactor pressure vessel. In each instance, this tube suspends from its upper end downwardly and slidingly through the suspension rod and casing to terminate with a lower end below the fuel rods and in the casing's lower end where, for example, the tube contains a thermocouple from which electrical lines extend upwardly through the instrumentation tube to the outside of the pressure vessel. The instrumentation tube is, in each instance, made so it can be removed upwardly by sliding upwardly through the fuel assembly casing and the tubular suspension rod suspension by which fuel assembly is suspended. The upper ends of the instrumentation tubes are fixedly supported by connections entirely separate from the connections by which the upper ends of the tubular suspension rods are fixed, although the upper ends of the instrumentation tubes must be releasable so that the tubes can be drawn upwardly, to assure against the hypothetical possibility that one or more of the instrumentation tubes might be inadvertently released for falling during operation of the reactor. The connections supporting the top ends of the instrumentation tubes are made with a redundancy of security or resistance to inadvertent release, and the instrumentation tubes are themselves made of metal and with dimensions providing a redundancy of tensile strength. SUMMARY OF THE INVENTION The present invention provides a secondary emergency suspension for each of the fuel elements without substantially altering the above described prior art construction, thereby eliminating any need for redesigning or reengineering that time-proven construction. In addition, this secondary or emergency suspension is provided by simple parts which are very inexpensive relative to other possible expedients. Keeping in mind that in the case of an accidentally released fuel assembly, both its instrumentation tube and all or at least most of the adjacent fuel assemblies can reasonably be expected to remain securely suspended and form fixed parts relative to the released assembly, this invention provides a simple latch means for normally latching each casing of each of the assemblies, to one of such parts that can be expected to remain securely suspended, thus providing for each of the fuel elements, a secondary or emergency suspension without the use of any extra parts other than those required for the latch means. With each assembly casing latched or locked to either its instrumentation tube or one or all of the casings of the assemblies adjacent to that assembly, the problem is presented of unlatching the latch means when it is desired to intentionally release the assembly for downward movement during core sevicing. At that time, the concrete pressure vessel in which the core is positioned, is designed to permit access upwardly to the bottoms of the fuel elements. With this in mind, the present invention provides each of the latch means with a latch release means on the inside of the casing, together with a tool that is insertable upwardly via the casing's open lower end, in each instance, for actuating the latch release means to release the latch means. The same tool can be used for one assembly after another or several of the tool means could be used simultaneously to release the latch means as to a group of the fuel assemblies. The tool means is, of course, used when the reactor is shut down and it is necessary to lower one or more of the fuel assemblies downwardly for core servicing operations. Even though the latch means is in simple form and does not apply vertically directed force to any of the parts, in the manner of a clamp means, and even though the latch means involves some looseness with respect to the interlatched parts, any looseness involved can be made so minor in the vertical direction as to catch an accidentally released fuel assembly almost immediately and before it has achieved any downward velocity sufficient to acquire destructive momentum with respect to the fixed parts to which it is latched. The above factors mean that if each assembly is latched to its instrumentation tube, the latter is able to reliably prevent the assembly from falling, or if it is latched to the casings of adjacent assemblies, their normal primary suspension means are well able to carry the weight of the released assembly. Correspondingly, the latch means may be made in the form of relatively small latches which do not interfere with the downward coolant gas flow through the core to an appreciable degree, while permitting the latch means to be provided by latches of simple design and which, therefore, do not add greatly to the overall cost of the core construction. The necessary tool means for releasing the latches is of simple construction and is, of course, capable of being used repeatedly. |
047388218 | summary | CROSS REFERENCE TO RELATED APPLICATIONS Reference is hereby made to the following copending applications dealing with related subject matter and assigned to the assignee of the present invention: 1. "Reusable Locking Tube in a Reconstitutable Fuel Assembly" by John M. Shallenberger et al., assigned U.S. Ser. No. 719,108 and filed Apr. 2, 1985 (W.E. 52,507). 2. "Improved Guide Thimble Captured Locking Tube in a Reconstitutable Fuel Assembly" by Robert K. Gjertsen et al., assigned U.S. Ser. No. 775,208 and filed Sept. 12, 1985 (W.E. 52,881). BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates generally to reconstitutable fuel assemblies for nuclear reactors and, more particularly, is concerned with a reconstitutable nuclear fuel assembly having a top-nozzle-to-guide-thimble attachment system employing reusable locking tubes with preformed dimples. 2. Description of the Prior Art In a typical nuclear reactor, the reactor core includes a large number of fuel assemblies each of which is composed of top and bottom nozzles with a plurality of elongated transversely spaced guide thimbles extending longitudinally between the nozzles and a plurality of transverse support grids axially spaced along and attached to the guide thimbles. Also, each fuel assembly is composed of a plurality of elongated fuel rods transversely spaced apart from one another and from the guide thimbles and supported by the transverse grids between the top and bottom nozzles. The fuel rods each contain fissile material and are grouped together in an array which is organized so as to provide a neutron flux in the core sufficient to support a high rate of nuclear fission. The reactor also has control rods which can be inserted into the guide thimbles to control the fission reaction. The fission reaction releases a large amount of energy in the form of heat. A liquid coolant is pumped upwardly through the core in order to extract some of the heat generated in the core for the production of useful work. During operation in the nuclear reactor, the fuel rods may occassionally develop cracks along their lengths resulting primarily from internal stresses. These defective fuel rods must be replaced in the fuel assemblies, and this replacement must occur under water as the fuel assemblies become highly radioactive during their operation in the reactor. To gain access to a defective fuel rod, it is necessary to remove the top and/or bottom nozzle of the fuel assembly. Reconstitutable fuel assemblies exist which are designed with removable nozzles. Typical removable top (bottom) nozzles have been attached to the top (bottom) of the guide thimbles using a threaded arrangement. Typical removable top nozzles also have been attached to the top of the guide thimbles using a bulge/groove arrangement, including the use of locking tubes, such as disclosed in U.S. Pat. No. 4,631,168, hereby incorporated by reference. Commonly owned U.S. patent application Ser. No. 719,108 entitled "Reusable Locking Tube in a Reconstitutable Fuel Assembly" by John M. Shallenberger et al., filed Apr. 2, 1985 (W.E. 52,507), is hereby incorporated by reference. The invention disclosed therein is a reconstitutable nuclear fuel assembly having reusable locking tubes with preformed dimples, and it has been in use and on sale in the U.S. for more than one year. Some locking tubes in those fuel assemblies were found to be seated too low. A study revealed that this problem had two possible causes. A properly seated locking tube could become improperly seated too low due to the force exerted on it when the fuel assembly was lifted, such as at the reactor site at the time of loading the fuel assembly into the reactor core. Also, a locking tube easily could be seated too low at the time of installation. A low-seated locking tube would require special (longer-handled) tooling for locking tube removal in the event the fuel assembly required reconstitution. This would complicate the underwater reconstitution operation. What is needed in a removable locking tube design which would insure proper seating of the locking tube at the time of installation and after fuel assembly lifting, and which would avoid time-consuming measurement checks at the factory and at the reactor. SUMMARY OF THE INVENTION It is an object of the invention to provide a reconstitutable nuclear fuel assembly having a top-nozzle-to-control-rod-guide-thimble attachment system employing reusable locking tubes of a design which insures proper locking tube seating at the time of installation. It is another object of the invention to provide such a locking tube design which also maintains proper locking tube seating during handling of the fuel assembly. Additional objects, advantages and novel features of the invention will be set forth in part in the description which follows, and in part will become apparent to those skilled in the art upon examination of the following or may be learned by practice of the invention. The objects and advantages of the invention may be realized and attained by means of the instrumentalities and combinations particularly pointed out in the appended claims. To achieve the foregoing and other objects and in accordance with the purpose of the present invention as embodied and broadly described herein, the reconstitutable nuclear reactor fuel assembly top-nozzle-to-control-rod-guide-thimble attachment system includes a top nozzle, a control rod guide thimble, and a locking tube. The top nozzle's adaptor plate has a control rod passageway. The passageway's upper portion has a smaller diameter than its lower portion, and is joined thereto by a ledge portion. The lower portion includes a circumferential groove. The thimble's top portion has more than one longitudinal, open-ended slot. Between the slots are fingers having a radially outwardly projecting rim or bulge. The rim has an inside surface defining a recess including a bottom edge. The thimble top portion is coaxially disposed in the passageway with the finger ends longitudinally placed proximate the ledge portion and with the rim (bulge) portions transversely placed in the groove. The locking tube includes an annular flexible cylinder which is coaxially placed in the thimble's top portion. The cylinder has at least one upper embossed dimple with an apex placed at an elevation below and proximate the ledge portion. The cylinder also has at least two lower embossed dimples each with a tip placed at an elevation above and proximate the elevation of the bottom edge of the recess. The upper dimples project radially outward into the passageway's lower portion at a distance, from the cylinder's centerline, which is greater than half the passageway's upper portion's diameter. The lower dimples project radially outward into the recess at a distance, from the cylinder's centerline, which is greater than the difference between the thickness of a finger and half the passageway's lower portion's diameter. The lower dimples are spaced angularly apart such that when at least one of the lower dimples is angularly rotated to be placed in a slot, at least one of the other lower dimples is angularly oriented to be aligned with the inside surface of the rim of one of the fingers. Another embodiment of the invention is the locking tube itself, the locking tube having been previously described within the context of the top-nozzle-to-control-rod-guide-thimble attachment system recounted in the previous paragraph. Several benefits and advantages are derived from the invention. The upper dimple feature prevents upward movement of the locking tube. The lower dimple feature prevents downward movement of the locking tube. The upper and lower dimple features together provide for proper seating of the locking tube at installation and provide for maintaining such proper seating during fuel assembly handling. |
042882907 | abstract | In apparatus for exchanging a control rod drive mechanism of a nuclear reactor of the type comprising a horizontal platform supported to be rotatable in a working chamber disposed below a reactor pressure vessel and a traveling carriage traveling on a rail assembly laid on the platform, there is provided a beam attached to the traveling carriage to be swingable between the vertical and horizontal positions and provided with a carrier for vertically moving the control rod drive mechanism. A holding device is secured to the beam so as to hold the control rod drive mechanism when the beam is moved to the horizontal position. The bolts which are used to connect the control rod drive mechanism to a housing in the pressure vessel are loosened and clamped by a bolt mounting device, which is conveyed in and out of the passage of movement of the carrier. There is further provided a carriage for receiving the control rod drive mechanism when the beam is moved to the horizontal position and for conveying the mechanism into an inspection chamber for inspecting the control rod drive mechanism. |
claims | 1. A method of direct conversion of nuclear energy to electrical energy comprising the steps of:placing a liquid semiconductor between two metal contacts, wherein said first metal contact creates a low resistance contact with said liquid semiconductor and said second metal contact creates a Schottky contact with said liquid semiconductor;placing nuclear material in solution in said liquid semiconductor;creating an electrical circuit between said first metal contact and said second metal contact, andgenerating electrical current and providing said current to said electrical circuit. 2. The method of claim 1, wherein the nuclear material comprises a radioactive isotope. 3. The method of claim 1, wherein the liquid semiconductor is a P-type semiconductor. 4. The method of claim 1, wherein the liquid semiconductor comprises selenium. 5. A method of conversion of nuclear energy to electrical energy comprising the steps of:providing a plurality of cells, said cell each comprising:a first metal contact layer having a first surface;a second metal contact layer having a first surface, said first and second metal contact layers positioned with said respective first surfaces in spaced relation to one another and facing each other to at least partially define a channel;a liquid semiconductor in said channel between said first metal contact layer and said second metal contact layer, wherein said first surface of said first metal contact layer forms a Schottky contact with said liquid semiconductor, and said first surface of said second metal contact layer forms a low resistance contact with said liquid semiconductor;a radioactive isotope in solution in said liquid semiconductor; andplacing an electrical load on said first and second metal contact layers of each cell in said plurality of cells, andgenerating electrical current and providing said current to said electrical load. 6. The method of claim 5, wherein said cells are electrically coupled together such that the outputs of the cells are combined. 7. The method of claim 5, wherein the liquid semiconductor is a P-type semiconductor. 8. The method of claim 5, wherein the liquid semiconductor comprises selenium. 9. The method of claim 5, wherein said radioactive isotope is dissolved in said liquid semiconductor. 10. The method of claim 2, wherein said radioactive isotope is dissolved in said liquid semiconductor. |
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claims | 1. A boiling water reactor nuclear power plant, in which a cooling water is circulated, comprising, in an installed state thereof: a reactor building; a reactor containment vessel positioned in the reactor building; said reactor containment vessel having dual cylindrical structure having inner and outer wall sections; a reactor pressure vessel disposed inside the containment vessel; a dry well defined, inside the reactor containment vessel, by the inner wall section thereof; a pressure suppression pool provided inside the reactor containment vessel and outside the dry well between the inner and outer wall sections of the reactor containment vessel; a containment vessel cooling system pool disposed above the suppression pool; a reactor core mounted with fuel assemblies supported by a reactor core support plate and an upper grid plate provided plate provided in an inner base portion of the reactor pressure vessel, said reactor core being disposed below said pressure suppression pool; a reactor core shroud surrounding the reactor core and the upper grid plate; control rod guide tubes positioned in the reactor core shroud and over the upper grid plate; control rods inserted in the control rod guide tubes; and control rod drive mechanisms operative for inserting and withdrawing the control rods from an upper portion of the reactor core, said control rod drive mechanisms being arranged at a portion above the control rod guide tubes and inside the reactor core shroud, said inner and outer wall sections of the reactor containment vessel having an inner hollow structure, the double-wall structure being communicated with the containment vessel cooling system pool, wherein said cooling water therein flows and circulates in the hollow portion of the double-wall structure to cool the dry well, the double-wall structure being provide with a plurality of ribs. 2. The boiling water reactor nuclear power plant according to claim 1 , wherein said pressure suppression pool being connected to said nuclear reactor pressure vessel by means of gravity-based piping through which the cooling water drops by gravity. claim 1 3. The boiling water reactor nuclear power plant according to claim 1 , wherein a piping and nozzles connected to said nuclear reactor pressure vessel are positioned above said reactor core. claim 1 4. The boiling water reactor nuclear power plant according to claim 1 , wherein a valve operable to open to an exterior of said reactor core shroud is provided at a position above said fuel assembly. claim 1 5. The boiling water reactor nuclear power plant according to claim 1 , wherein said pressure suppression pool and a lower portion of the dry well are connected by means of a plurality of emergency opening passages at different elevational positions. claim 1 6. The boiling water reactor nuclear power plant according to claim 1 , wherein a normal use cooling system is connected to the inner hollow structure of the reactor containment vessel wall. claim 1 7. The boiling water reactor nuclear power plant according to claim 1 , wherein a normally-closed water discharge pipe is led from said pressure suppression pool into said dry well at a base region of said nuclear reactor pressure vessel, and said water discharge pipe is normally closed by a sealing device capable of being released in case of emergency so as to open said water discharge pipe. claim 1 8. The boiling water reactor nuclear power plant according to claim 1 , wherein a heat pipe capable of exchanging heat is provided at a portion between said pressure suppression pool and a lower region of said dry well. claim 1 9. The boiling water reactor nuclear power plant according to claim 1 , wherein a guard pipe is provided so as to extend from said dry well section to said pressure suppression pool, and valves and piping led from said nuclear reactor pressure vessel are accommodated in said guard pipe. claim 1 10. The boiling water reactor nuclear power plant according to claim 1 , wherein a turbine system is installed on an upper portion of the reactor building. claim 1 11. The boiling water reactor nuclear power plant according to claim 1 , wherein an extraction space capable of accommodating said nuclear reactor pressure vessel is provided above the nuclear reactor pressure vessel in the reactor building. claim 1 12. The boiling water reactor nuclear power plant according to claim 1 , wherein said reactor building is positioned on a foundation base having an anti-seismic structure. claim 1 13. A reactor containment vessel for use with a boiling water nuclear reactor having a reactor containment vessel cooling system providing cooling water, comprising: an inner wall made from multiple steel plates defining an inside of the reactor containment vessel; an outer wall made from multiple steel plates, wherein the inner wall and the outer wall are positioned to form a double-wall structure forming an inner hollow structure over at least a portion of the reactor containment vessel; a plurality of ribs provided within the inner hollow structure and coupled to either or both of the inner wall and the outer wall and; a fluidic connection to the reactor containment vessel cooling system configured so that cooling water from the reactor containment vessel cooling system flows and circulates in the inner hollow structure to cool a portion of the inside of the reactor containment vessel. |
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description | FIG. 1 is a sectional view, with parts cut away, of a boiling water reactor (BWR) 8 including a reactor pressure vessel (RPV) 10. RPV 10 has a generally cylindrical shape and is closed at one end by a bottom head 12 and at its other end by a removable top head 14. A side wall 16 extends from bottom head 12 to top head 14. A cylindrically shaped core shroud 20 surrounds a reactor core 22. Shroud 20 is supported at one end by a shroud support 24 and includes a removable shroud head 26 at the other end. An annulus 28 is formed between shroud 20 and side wall 16. Heat is generated within core 22, which includes fuel assemblies 36 of fissionable material. Water circulated up through core 22 is at least partially converted to steam. Steam separators 38 separates steam from water, which is recirculated. Residual water is removed from the steam by steam dryers 40. The steam exits RPV 10 through a steam outlet 42 near vessel top head 14. Fuel assemblies 36 are aligned by a core plate assembly 50 located at the base of core 22. A top guide 52 aligns fuel assemblies 36 as they are lowered into core 22. Core plate 50 and top guide 52 are supported by core shroud 20. Core spray supply pipes 54 supply coolant to the core 22 during a loss of coolant accident. FIG. 2 is a top view schematic of a core spray sparger assembly 70 positioned above top guide 52, shown in FIG. 1. Top guide 52 is a latticed structure including several top guide beams 72 defining top guide openings 74. Top guide openings 74 are sized to receive fuel assemblies 36. Core spray sparger assembly 70 includes coolant manifolds 80, coolant couplings 82 configured to mate with coolant supply pipes 54 (shown in FIG. 1), mounting devices 84 coupling coolant manifold 80 to BWR 8 (shown in FIG. 1), and fluid conductors 86 in a parallel array 88. Core spray sparger assembly 70 further includes nozzles 90 in fluid communication with fluid conductors 86. FIG. 3 is a perspective view of core spray sparger assembly 70. In one embodiment, two fluidically independent, redundant coolant manifolds 80, 92 are provided in a coaxial, substantially circular arrangement. Coolant manifold 80 includes an upper surface 94, a lower surface 96, an outer face 98 and an inner face 100, while coolant manifold 92 includes an upper surface 102, a lower surface 104, an outer face 106 and an inner face 108. In the exemplary embodiment, coolant manifolds 80 and 92 are concentric. In another embodiment, coolant manifolds 80 and 92 are stacked vertically (not shown). Coolant manifolds 80 and 92 are shown in FIG. 3 with substantially rectangular cross-sections, but other configurations include, for example, circular, square and oval cross-sections. Coolant manifolds 80 and 92 are joined together by resilient couplings 110. Resilient couplings 110 secure manifolds 80 and 92 together while facilitating differential thermal expansion between manifolds 80 and 92. In one embodiment, resilient coupling 110 includes a welded, metallic, U-shaped coupling 112 extending between upper surfaces 94 and 102, and lower surfaces 96 and 104. Coolant manifolds 80 and 92 are fluidically independent and redundant. Each coolant coupling 82 is fluidically coupled to one coolant manifold 80 or 92. Each fluid conductors 86 is fluidically coupled to one coolant manifold 80 or 92. A failure of any component connected to coolant manifold 80 does not prevent coolant manifold 92 from supplying coolant to fuel assemblies 36 (shown in FIGS. 1 and 2). In one embodiment, coolant manifolds 80 and 92 are each unitary constructs. In another embodiment, coolant manifolds 80 and 92 are formed using a plurality of sections. Core spray sparger assembly 70 also includes alignment guides 118 configured to align coolant manifolds 80 with coolant supply pipes 54 (shown in FIG. 1). In one embodiment, alignment brackets 120 extend radially from coolant manifold outer face 98. Each alignment bracket 120 is configured to be received in an alignment slot (not shown) in shroud head 26. Alignment guides 118 also align core spray sparger assembly 70 to top guide 52. In one embodiment, alignment cones 124 extend from coolant manifold lower surface 96, coolant manifold lower surface 104, or both, to engage alignment sockets (not shown) in top guide 52 (shown in FIGS. 1 and 2). Coolant couplings 82 join coolant supply pipes 54 (shown in FIG. 1) to coolant manifolds 80 and 92. The orientation, precise circumferential spacing, and number of coolant couplings 82 varies with specific reactor design considerations, but multiple, spaced coolant couplings 82 for each coolant manifold 80 and 92 facilitate the desired coolant flow volume and safety redundancy. In one embodiment, coolant couplings 82 include slip couplings with spherical seats (not shown), sized to receive supply pipes 54. As shown in FIG. 3, four coolant couplings 82 are welded to each coolant manifold lower surface 96 and 104 to facilitate supplying coolant from supply pipes 54 to each coolant manifold 80 and 92. In one embodiment, each lower surface 96 and 104 of coolant manifolds 80 and 92 are joined to a boron coolant coupling 130, configured to supply borated coolant. Coolant coupling 130 joins a selected coolant supply pipe (not shown) which is further connected to a borated coolant system (not shown). Mounting devices 84 facilitate retention of core spray sparger assembly 70 within RPV 10. Mounting devices 84 include hanger bolts 140 that secure coolant manifold 80 and 92 to shroud head 26. Each hanger bolt 140 includes a trunnion 142 and a pair of stanchions 144. Stanchions 144 are welded to coolant manifold 80 and 92. In one embodiment, three, symmetrically-spaced hanger bolts 140 mount to coolant manifold 80, secured to upper surface 94 and to outer face 98 and three hanger bolts 140 mount to coolant manifold 92, secured to upper surface 102 and inner face 108. In another embodiment, core spray sparger assembly 70 includes more than three or less than three hanger bolts 140. Hanger bolts 140 support core spray sparger assembly 70 against fluid forces and flow induced vibrations, while accommodating differential thermal expansion. Hanger bolts 140 and trunnions 142 facilitate radial differential thermal expansion between shroud head 26 and core spray sparger assembly 70 by allowing fractional rotation of hanger bolts 140 about trunnions 142. In one embodiment, mounting devices 84 support core spray sparger assembly 70 above top guide 52. In another embodiment, both mounting devices 84 and top guide 52 support core spray sparger assembly 70. Fluid conductors 86 are fluidically coupled to coolant manifolds 80 and 92 to form parallel array 88. Each fluid conductor 86 includes a longitudinal section 150, a proximate end 154, and a distal end 156. A connection section 152 at each proximate end 154 and each distal end 156 joins fluid conductors 86 to coolant manifolds 80 and 92 while facilitating differential thermal expansion. In one embodiment, connection sections 152 include transition elbows 158, which facilitate accommodation of thermal expansion and contraction. Fluid conductors 86 are separated into a first parallel array 160 and a second parallel array 162, fluidically independent of each other. Parallel arrays 160 and 162 are positioned above reactor top guide 52, with parallel array 160 in fluid communication with coolant manifold 80 and parallel array 162 in fluid communication with coolant manifold 92. First parallel array 160 and second parallel array 162 are interspersed with each other, such that fluid conductors 86 of first parallel array 160 alternate with fluid conductors 86 of second parallel array 162. Fluid conductors 86 of first array 160 are a horizontal spacing distance 170 from adjoining fluid conductors 86 of second array 162. In one embodiment, spacing distance 170 is substantially similar to the width of top guide opening 74, facilitating inspections and passage of fuel assemblies 36. First parallel array 160 and second parallel array 162 are configured such that each array 160 and 162 is proximate to each fuel assembly 36. Each array 160 and 162 is configured to supply coolant to each fuel assembly 36, providing redundant coolant flow. As shown in FIG. 3, fluid conductors 86 define a cylindrical cross-section. In alternate embodiments, fluid conductors 86 include rectangular, square or oval cross-sections. In one embodiment, fluid conductors 86 are stabilized against flow induced vibrations by a stabilizing member 180. Stabilizing member 180 includes attachment devices 182 securing fluid conductors 86 to stabilizing member 180. Stabilizing member 180 is coupled to at least one of coolant manifold 80 and 92, and further coupled to at least one fluid conductor 86. Attachment devices 182 include support clips 184 extending from stabilizing member 180 and welded to fluid conductors 86. In another embodiment, stabilizing member 180 is coupled to selected fluid conductor 86 by restraining cavities (not shown) formed in stabilizing member 180, each sized to receive and restrain one fluid conductor 86. Because fluid conductors 86 can vary in length, in one embodiment, only selected fluid conductors 86 are secured to stabilizing member 180. Nozzles 90 are fluidically coupled to each fluid conductor 86. Nozzles 90 are formed in each fluid conductor 86 by precision drilling, electric discharge machining (EDM), or other suitable techniques. In one embodiment, nozzle ports (not shown) are secured to each fluid conductor 86. More specifically, the nozzle ports are welded to fluid conductors 86. In another embodiment, nozzle ports are screwed into fluid conductors 86. Nozzle 90 location on each fluid conductor 86 is predetermined prior to forming or welding. Nozzles 90 are formed, including location and shape, in each fluid conductor 86 such that each array 160 and 162 supplies coolant to each fuel assembly 36. Each nozzle 90 is targeted to a specific fuel assembly 36 (shown in FIGS. 1 and 2). Each fuel assembly 36 receives coolant from one nozzle 90 in array 160 and from one nozzle 90 in array 162. In another embodiment, one nozzle 90 is targeted to provide coolant to more than one fuel assembly 36. In fabrication, nozzles 90 are formed in predetermined positions in fluid conductors 86. Fluid conductors 86 are then positioned in parallel arrays 160 and 162, a predetermined spacing distance 170 apart, aligned relative to each other, and joined to coolant manifolds 80 and 92. In another embodiment, parallel arrays 160 and 162 are joined to coolant manifolds 80 and 92, and then nozzles 90 are formed in predetermined location in fluid conductors 86. Spacing distance 170 between adjacent fluid conductors 86 is generally coordinated with top guide 52 (shown in FIG. 1) to facilitate minimizing obstruction of flow pass sparger assembly 70. In one embodiment, spacing distance 170 facilitates removal of fuel assemblies 36 during maintenance. Parallel arrays 160 and 162 are positioned above top guide beams 72 (shown in FIG. 2). Parallel array 160 and 162 are aligned to fuel assemblies 36 such that each fuel assembly 36 is targeted to receive coolant. More specifically, each parallel array 160 and 162 is configured such that each fuel assembly 36 receives coolant from each array 160 and 162. In use, coolant couplings 82 are configured to receive supply pipes 54 and are in fluid communication with coolant manifolds 80 and 92. Fluid conductors 86 are in fluid communication with coolant manifold 80 such that coolant flows from the supply pipes 54, through coolant couplings 82, through coolant manifold 80 to fluid conductors 86. Coolant in fluid conductors 86 flows to nozzles 90. Parallel array 160 and 162 are aligned to fuel assemblies 36 and configured such that each fuel assembly 36 is targeted to receive coolant from each array 160 and 162 when coolant is supplied to core spray sparger assembly 70. FIG. 4 is a perspective, sectional, view of another embodiment of a core spray sparger assembly 300. Core spray sparger assembly 300 includes a pair of coolant manifolds 302 and 304, coolant coupling 306, mounting devices 308, fluid conductors 310 and 312, and nozzles 314. Coolant manifolds 302, coolant coupling 306, and nozzles 314 are substantially identical, respectively, to coolant manifolds 80, coolant coupling 82, and nozzles 90 of core spray sparger assembly 70 described above. Mounting devices 308 include leaf spring retainers 316 configured to secure core spray sparger assembly 300 between top guide 52 and shroud head 26 (shown in FIG. 1). Leaf spring retainers 316 are secured to each coolant manifold 302 and 304, such that leaf spring retainers 316 engage shroud head 26 when core spray sparger assembly 300 and shroud head 26 are installed in RPV 10. Leaf spring retainers 316 facilitate radial and vertical thermal expansion of core spray sparger assembly 300. In another embodiment, core spray sparger assembly 300 includes more than three or less than three leaf spring retainers 316. Fluid conductors 310 include alignment guides 320. Alignment guides 320 are located on fluid conductors 310 to align core spray sparger assembly 300 to top guide 52 (shown in FIGS. 1 and 2). More specifically, fluid conductors 310 include channels 320 that mate to corresponding tongues (not shown) on top guide 52. Channels 320 facilitate aligning fluid conductors 310 such that nozzles 312 are aligned to supply coolant to each fuel assembly 36. Channel 320 may extend for less than the length of fluid conductor 310. In one embodiment, channels 320 substantially receive top guide beams 72. In one embodiment, only selected fluid conductors 310 include channels 320. In another embodiment, top guide 52 can include channels while fluid conductors 310 include corresponding tongues. Fluid conductors 312 are positioned above fluid conductors 310. Fluid conductors 312 include a longitudinal section 324 and slip seats 326 that mate to coolant manifold 304. Slip seats 326, in conjunction with slip couplings 328 in an inner face 330 of coolant manifold 304, facilitate thermal expansion and contraction of fluid conductors 312. Coolant manifold 304 is substantially identical to coolant manifold 92, with the exception of receiving fluid conductors 312 in slip couplings 328 in inner face 330. Two fluid conductors 312 are mated to coolant manifold 304 positioned above fluid conductors 310. Fluid conductors 312 are in a first array 332, which is in flow communication with coolant manifold 304. First array 332 includes all fluid conductors 312 and 310 in flow communication with coolant manifold 304. A second array 334 includes all fluid conductors 310 in flow communication with coolant manifold 302. Fluid conductors 310 and 312 are in either first array 332 or second array 334. In one embodiment, all fluid conductors in first array 332 are configured to mate with coolant manifold 304 through slip coupling seat 326 in inner face 330. Core spray sparger assembly 70 facilitates distribution of coolant to each fuel assembly 36, minimizes the requirement for field alignment, and minimizes in-vessel inspection complexities to improve reactor maintenance practices and provide for great efficiency. While the invention has been described in terms of various specific embodiments, those skilled in the art will recognize that the invention can be practiced with modification within the spirit and scope of the claims. |
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052020843 | description | MODE(S) FOR CARRYING OUT THE INVENTION Illustrated schematically in FIG. 1 is a nuclear reactor 10 in accordance with an exemplary embodiment of the present invention. The reactor 10 includes an annular reactor pressure vessel 12 having a vertical, or longitudinal centerline axis 14. The vessel 12 contains a reactor coolant 16, such as water, filling the vessel 12 to a preselected level L near the upper middle portion of the vessel 12 between its top and bottom heads. An annular core shroud 18 extends coaxially about the centerline axis 14 and is spaced radially inwardly from the vessel 12 to define an annular downcomer 20. The core shroud 18 is also spaced upwardly from the bottom end of the vessel 12 to define a conventional lower plenum 22 disposed in flow communication with the downcomer 20. A first, or lower, reactor core 24 is disposed at the bottom of the core shroud 18 in flow communication with the lower plenum 22. The lower core 24 includes a plurality of conventional first, or lower, fuel bundles 26 configured in a conventional first two-dimensional array or matrix with adjacent fuel bundles 26 being laterally, or radially, spaced apart as is conventionally known. The lower core 24 operates as a boiling water reactor (BWR) which heats and boils the water 16f received from the lower plenum 22 to form a steam and water mixture 16a which rises upwardly through the lower core 24. A plurality of laterally spaced apart steam separators 28 are disposed above the lower core 24 in flow communication therewith for receiving the steam and water mixture 16a, and are effective for separating the water as liquid, i.e. separated water 16b, from the steam, i.e. separated steam 16c. A second, or upper reactor core 30 is disposed within the core shroud 18 above the steam separators 28 in flow communication therewith for receiving and heating the separated steam 16c to form superheated steam 16d. The upper core 30 heats the separated steam 16c which thereby cools the upper core 30 thusly forming a steam cooled reactor (SCR). The upper core 30 includes a plurality of conventional second, or upper fuel bundles 32 similarly conventionally configured in a second two-dimensional array or matrix. In this exemplary embodiment, the core shroud 18 is used and defines the downcomer 20. However, in alternate embodiments, the core shroud 18 may be eliminated, and therefore, the first core 24, the steam separators 28, and the second core 30 are spaced radially inwardly from the vessel 12 to define the downcomer 20. The superheated steam 16d flows upwardly from the upper core 30 into a conventional steam dryer 34 which removes any remaining moisture therefrom and is then discharged from the reactor 10 through a conventional outlet nozzle 36. The reactor 10 also includes a conventional annular feedwater sparger 38 disposed above the lower core 24 and in flow communication with the downcomer 20 for conventionally channeling into the downcomer 20 relatively cool feedwater 16e piped from an inlet nozzle 39 which mixes with the water 16 in the downcomer 20, flows downwardly into the lower plenum 22 and then flows into the lower core 24 as inlet water 16f. The water 16 is naturally circulated within the vessel 12 by the difference in density of the water 16, with its density being greatest in the downcomer 20 due to its relatively low temperature, and its density being relatively low in the lower core 30 since it is relatively hot and mixed with steam in the steam and water mixture 16a. Optionally, conventional pumping means such as jet pumps or reactor internal pumps (not shown) can be used to further assist recirculation. The lower and upper fuel bundles 26, 32 preferably share a common form, i.e. substantially identical in configuration so that each fuel bundle can be placed in any position in either matrix. During refueling operations net transfers are as follows: spent bundles are removed from the lower core 24, partially spent bundles from the upper core 30 are inserted into the lower core 24, and fresh bundles are inserted into the upper core 30. This fuel bundle shuffling is an average refueling scheme and does not exclude the possibilities that some bundles are retired from the upper core 30, some fresh fuel bundles are inserted into the lower core 24, and that some partially spent fuel bundles are transferred from the lower core 24 to the upper core 30. Due to heating by the cores 24 and 30, and due to water/steam separation by the separators 28, the steam void fraction of the water 16 increases at higher levels in the vessel 12 so that the steam void fraction is greatest at the top of the upper core 30 than it is at the bottom of the lower core 24. Accordingly, neutron moderation is more effective at the lower level than at the upper level. Because of the difference in moderation, fuel bundles 32 in the upper core 30 are subjected to a harder neutron spectrum than are the fuel bundles 26 in the lower core 24. The harder neutron spectrum can be taken advantage of by the fresh fuel bundles in the upper core 30. The harder neutron spectrum contains a higher percentage of fast and epithermal neutrons, while the thermal neutron spectrum contains a higher percentage of slower thermal neutrons. Thermal neutrons are more effective than faster neutrons at causing fission. The faster neutrons are more likely to be subjected to capture or resonance absorption reactions which do not result in fission. Non-fissioning neutron absorption results in isotopic enhancement. In other words, the hard neutron spectrum converts fertile material to fissile fuel. The primary reaction is the absorption of a fast neutron by fertile U.sup.238 to yield fissile Pu.sup.239 through a relatively short-lived radioactive decay chain. Neutron absorption by Pu.sup.239 can result in fission or in the formation of the next plutonium isotope, fertile Pu.sup.240. Neutron absorption by fertile Pu.sup.240 results in a fissile Pu.sup.241 isotope. The net effect of the hard neutron spectrum is production of additional fissile material as the original fissile material is partially spent. Furthermore, other transuranics which are formed through various nuclear reactions and radioactive decay chains have a good likelihood to fission in the hard spectrum, producing useful energy and minimizing high level, actinide-series waste products. Thus, the relatively hard neutron spectrum of the upper fuel bundles 32 can be used to convert fertile fuel to fissile fuel and minimize waste poisons, enhancing the operational lifetime of a fuel bundle. The harder neutron spectrum in the upper core 30 is less effective in inducing fission. This is not a problem where relatively fresh fuel bundles in the upper core 30 are fuel enriched to contain relatively high concentrations of fissile fuel, for example, U.sup.235. As the U.sup.235 is depleted faster than additional fissile fuel is created, the hard neutron spectrum would eventually be unable to support a chain reaction. Prior to this point, the no-longer-fresh fuel bundle can be transferred from the upper core 30 to the lower core 24, which is exposed to a more thermal neutron spectrum. Since thermal neutrons are most effective at inducing fission, fuel in the lower core 24 can be more fully utilized. This provides advantages in fuel economics as well as waste disposal. Since the fuel in the lower core 24 is subjected to a thermal spectrum, the ratio of capture and resonance absorption to fission reactions is less, resulting in less conversion of fertile materials and less high-level waste. Thus, isotopic enhancement, which might otherwise contribute to higher levels of long lived radioactivity in the spent fuel elements due to plutonium production, is minimized in the soft neutron spectrum of the lower core 24. The present invention provides for enhanced fuel arrangement flexibility which can take advantage of axial neutron spectral shifts through the core. As a result, fuel lifetimes are increased and the quantity of high-level nuclear waste is minimized. These and other features and advantages of the present invention are apparent in the following detailed description. The bi-level fuel bundle arrangement of the present invention provides additional flexibility in the redistribution of fuel bundles during refueling operations. In particular, an axial level, in addition to the conventional radial array position, can be selected for each fuel bundle. This provides for a refueling scheme in which fresh or low burnup fuel bundles are installed in upper core 30 where a harder neutron spectrum can convert fertile fuel to fissile fuel. Partially spent, or medium to high burnup, fuel bundles can be moved from upper core 30 to the lower core 24 where the more thermal, or soft, neutron spectrum can more effectively utilize, or fission, the remaining fissile fuel. Fertile fuel conversion is minimized in the lower core 24 due to the soft neutron spectrum so that a relatively complete burnup is possible while minimizing the quantity of high-level radioactive waste products in the lower fuel bundles 26. Fuel bundles in a conventional one-level core are typically subject to several burn cycles before being disposed. At each refueling outage, the bundles are redistributed radially in the matrix with those having less burnup being moved to alternate radial positions for obtaining more complete burnup in view of the radial variation in neutron flux density and spectral distribution. Once a fuel bundle is substantially burned, i.e. has undergone substantially complete fission, it is then retired from the core. Accordingly, the present invention allows for additional flexibility in fuel management by allowing axial, as well as radial, redistribution of fuel bundles. Fresh, or low burnup fuel is preferably provided in the upper core 30, and medium to high burnup fuel is preferably provided in the lower core 24. The low burnup fuel for the upper core 30 may come from radial redistribution within the upper core 30 itself, and the medium or high burnup fuel for the lower core 24 preferably comes from the upper core 30, with a portion thereof coming from radial redistribution of the lower core 24 as desired. The present invention provides for cores with two or more levels of fuel bundles, which can be monolithic or contain multiple elements. The bundles on a level can be packed in two-dimensional arrays, or matrices, as triangles, rectangles including squares, or hexagons. Other packing shapes may also be used. Some embodiments employing control rods need not use them at all core levels. Power output regulation can be conventionally effected using burnable poisons, adjusting coolant flow and temperature, and/or using other power regulation approaches. Access to the lower core 24 can be through the upper core 30, from the bottom, or through lateral access. The lower core 24 is operated as a conventional boiling water reactor which converts the subcooled inlet water 16f to a predeterminedly high-void fraction mixture 16a at about 7.0 MPa and about 271.degree. C. The upper core 30 is fed by the separated steam 16c from the separators 28 and converts it to the superheated steam 16d, for example, at about 7.0 MPa and temperatures greater than about 271.degree. C. The fuel bundle configuration or mechanical design for the two stages is preferably identical, and the upper core 30 is loaded with fresh or low burnup fuel while the lower core 24 is loaded with fuel that has undergone at least one cycle of exposure in the upper core 30 as described above. After complete burnup operation in the lower core 24, the fuel is discharged. Accordingly, the lower core 24 provides a compact steam input source for the steam cooled upper core 30 without the need for complex and inefficient steam blowers and injectors to achieve steam circulation as found in conventional steam cooled reactor concepts. Neither is there a need for conventional contact boilers to provide the steam source. The movement of fuel from the upper to the lower cores has several advantages. It initially maximizes in the upper, or converter, core 30 the conversion ratio of fertile U.sup.238 and Pu.sup.240 by resonance capture in the hard neutron spectrum, isotopically enhancing the fertile fuel to fissile fuel Pu.sup.239 and Pu.sup.241. Subsequently, it maximizes in the lower, or burner, core 24 burnout of the Pu in the thermal spectrum of the boiling stage, while limiting conversion of U.sup.238 near the end of bundle life. Furthermore, maximum flexibility is provided for achieving power distribution shaping by means of increased fuel enrichment in the upper core 30, which is not wasted since remaining fuel from the upper core 30 is subsequently burned in the lower core 24, and by means of burnable poisons in the lower core 24 to reduce burnup in the lower core 24 to better match the burnup in the upper core 30 for obtaining a more uniform axial burnup. With uranium fuel, the upper core 30 effects relatively high conversion with a nuclear lifetime of about 150,000 megawatt day per metric tonne (MWD/Tonne) being attainable. With plutonium fuel, the upper core 30 can be designed to breed fuel, with a nuclear lifetime of about 200,000 MWD/Tonne being attainable. In either case, high conversion is achieved and significant reductions in the generation of high-level waste is realized by first burning in the upper core 30, then subsequently burning in the lower core 24. This offers the advantage of recycling the converted fertile fuel from the upper core 30 to the lower core 24, and burnup in the lower core 30 of significant quantities of remaining high level wastes which did not fission during its residence in the upper core 30, without having to conventionally chemically reprocess the fuel. It should be appreciated that the reactor 10 has a neutron spectrum from the bottom of the lower core 24 to the top of the upper core 30 which is varying from relatively soft to hard. As indicated above, axial variation in both neutron flux density and neutron spectrum occurs from boiling the water 16 and creating void fractions. Since the steam 16c is channeled upwardly through the upper core 30 and undergoes superheating, the neutron spectrum therein is relatively hard and substantially harder than that found in the lower core 24. Accordingly, the reactor 10 may take advantage of this axial variation in neutron spectrum with the resulting hard neutron spectrum in the upper core 30, which is substantially greater than that found in a conventional BWR to provide fertile fuel conversion to fissile fuel. The upper core 30 may then be considered a converter stage, and as mentioned above, with plutonium used as a fuel, the upper core 30 can be designed to breed fuel, i.e., generate more usable fuel than it consumes. Accordingly, the upper fuel bundles 32 preferably include fertile fuel which forms fissile fuel by conversion due to the hard neutron spectrum. And, the lower fuel bundles 26 preferably include fissile fuel which is fissionable due to the soft neutron spectrum found in the lower core 24. In order to more easily redistribute the upper fuel bundles 32 into the lower core 24 for use as the lower fuel bundles 26, the upper and lower fuel bundles 32 and 26 are preferably identical in configuration having the same height and lattice arrangement. And, each of the steam separators 28 is preferably vertically aligned with and interposed between respective ones of the upper and lower fuel bundles 32, 36, and may be removed therewith as a single three-component assembly, if desired. More specifically, illustrated in FIG. 2 is an exemplary configuration of the lower and upper fuel bundles 26 and 32 in a hexagonal lattice. The lower and upper fuel rods thereof are preferably arranged in a hexagonal configuration and spatially joined together in the transverse, or radial plane, by a pair of axially spaced apart tie plates 40. Each of the tie plates 40 includes a respective aperture 42 through which a respective fuel rod is positioned, along with additional flow apertures 44 through which the inlet water 16f, mixture 16a, separated steam 16c, and superheated steam 16d flow, respectively. Accordingly, each of the lower and upper fuel bundles 26 and 32 are preferably identical and have an identical radial configuration, e.g., hexagonal, and identical longitudinal length L.sub.1 and L.sub.2 so that individual ones of the upper fuel bundles 32 may be simply axially redistributed into the lower core 24 to replace one of the lower fuel bundles 26. This is a significant advantage of the present invention since the preburned upper fuel bundles 32 may be moved from the upper core 30, subject to the hard neutron spectrum, to the lower core 24, subject to the soft neutron spectrum, wherein they may be more fully and completely burned. In this way, when the upper fuel bundles 32 are initially positioned in the upper core 30 they may be provided with more fuel enrichment for obtaining a more uniform axial power distribution from the entire reactor 10 without wasting such enriched fuel as would occur in a conventional single level reactor wherein such enriched fuel could not practically be reutilized. But, in accordance with the present invention, the upper fuel bundles 32 may be recycled into the lower core 24 for more complete burning. Furthermore, in an alternate embodiment where the fuel pellets from the lower and upper fuel bundles 26, 32 are reprocessed for separating and collecting substantial amounts of actinide-series elements, the collected actinide-series elements may be formed into pellets along with other fissile and fertile fuel makeup to be preferably placed in the upper fuel bundles 32 for being exposed to the hard neutron flux to fission and produce useful energy while further decreasing the high level wastes associated therewith. As illustrated in FIG. 2, the lower fuel bundles 26 may be conventionally simply supported on a lower stationary support plate 46 at the bottom of the lower core 24, which support plate 46 has a plurality of flow apertures 48 extending axially therethrough. These flow apertures 48 allow the inlet water 16f to circulate upwardly through the lower core 24 and between adjacent fuel bundles and within individual fuel bundles as is conventionally known. Since water, or a water/steam mixture is not being channeled past the upper fuel bundles 32, conventionally required flow channels or baffles are not required, and, therefore, the upper fuel rods are disposed directly in contact with the separated steam 16c which is allowed to flow freely between adjacent fuel rods of adjacent upper fuel bundles 32. However, since the water 16a flows upwardly over the lower fuel bundles 26, suitable flow channels or baffles are still required for controlling the flow thereof. More specifically, in one embodiment of the present invention each of the lower fuel bundles 26 is disposed in an annular flow baffle 50 for its entire axial length L.sub.1. The baffle 50 is hexagonal to match the hexagonal fuel bundles and channel therethrough and over the fuel rods the inlet water 16f for forming the steam and water mixture 16a. Each of the flow baffles 50 may be permanently attached to the lower support plate 46 so that the lower and upper fuel bundles 26 and 32 may remain identical in configuration, and, upon axial redistribution thereof, one of the lower fuel bundles 26 may be removed from within its baffle 50 and replaced by one of the upper fuel bundles 32. The tie plates 40 are preferably complementary in shape, i.e., hexagonal, so that they fit within the baffles 50. Alternatively, the baffles 50 may be provided as separate components in which the lower fuel bundles 26 are placed prior to being positioned in the lower core 24 above the support plate 46. Adjacent ones of the baffles 50 are spaced from each other to define a longitudinally extending bypass channel 52 for receiving and channeling downwardly the separated water 16b from the steam separators 28. Each of the baffles 50 extends upwardly above the first fuel bundle 26 to define an annular skirt portion, or simply skirt 54 spaced around a respective one of the steam separators 28 to define therewith a skimmer discharge passage 56. Since the skirt 54 is a portion of the baffle 50 it too is hexagonal in this exemplary embodiment. Each of the skirts 54 includes a plurality of circumferentially spaced bypass inlets 58 for channeling the separated water 16b from the steam separator 28 downwardly into the bypass channel 52. As shown in more particularity in FIG. 3, each of the steam separators 28 includes a tubular separating barrel 60 spaced radially inwardly from the skirt 54 to define the skimmer discharge passage 56. The separator 28 includes a first, imperforate nosepiece 62 generally of a truncated, tubular cone shape which defines an inlet 64 at the bottom of the separator 28 disposed in flow communication with a respective one of the first fuel bundles 26 for receiving the steam and water mixture 16a therefrom. The first nosepiece 62 is suitably sealingly joined to the upper tie plate 40 of the first fuel bundle 26 for receiving the mixture 16a from the flow holes 44, and converges upwardly and sealingly joins the separator inlet 64. The outer surface of the first nosepiece 62 forms a boundary for the lower part of the skimmer discharge passage 56 adjacent to the bypass inlets 58. The steam and water mixture 16a is channeled upwardly into the barrel 60 through conventional swirl vanes 66 which swirl the mixture 16a radially outwardly for allowing centrifugal force to separate liquid from the steam, with the separated steam 16c being discharged from the top of the separator 28 through a central steam outlet 68. The steam outlet 68 in this exemplary embodiment is a tubular member fixedly mounted to an annular support plate 70, which in turn is fixedly mounted inside the skirt 54. A second conical nosepiece 72 is fixedly joined in flow communication to the steam outlet 68 through a central aperture in the support plate 70, and diverges upwardly and is fixedly and sealingly joined to the lower tie plate 40 of the upper fuel bundle 32. The separated steam 16c flows upwardly from the outlet 68, through the second nosepiece 72 and through the flow holes 44 of the lower tie plate 40 for flow upwardly into the upper fuel bundle 32. An annular skimmer tube 73 extends downwardly from the steam outlet 68 inside the barrel 60. An annular skimmer liquid outlet 74 is disposed at the top of the separator barrel 60 and is in the form of a plurality of circumferentially spaced apertures disposed radially outwardly of the skimmer tube 73. Accordingly, as the mixture 16a is swirled by the swirl vanes 66 inside the barrel 60 the liquid is centrifuged radially outwardly toward the inner surface of the barrel 60 and flows upwardly between the skimmer tube 73 and barrel 60 and out of the liquid outlet 74 into the top of the skimmer discharge passage 56. The support plate 70 prevents the separated water 16b from flowing upwardly, which is instead, turned downwardly to flow through the skimmer discharge passage 56 and into the bypass inlets 58 for flow downwardly in the bypass channel 52. The steam separators 28 effectively use the skirt 54, which is a portion of the baffle 50, in separating liquid from steam. Although this exemplary embodiment of the steam separator 28 is a single stage separator, multi axial stage separators could also be used. Since the water level L in the vessel 12 is below the upper core 30 and near the top of the steam separators 28, the baffle 50 need not extend upwardly to the top of the upper core 30 since its flow channeling function is not required. However, it may be extended to the top of the upper core 30 for providing a convenient passage for guiding the lower and upper fuel bundles 26 and 32 into their respective cores during refueling. In a preferred embodiment, the baffles 50, and more specifically the skirts 54, extend upwardly into the upper core 30 to a relatively short predetermined height H from the lower support plate 46 for maintaining a predetermined level of the water 16 in the bypass channels 52 to provide a suitable differential head for promoting separation by the separators 28 of the steam 16c from the steam and water mixture 16a. The level of the water 16 in the bypass channels 52 may be selected for preventing excessive carryunder of steam in the separated water 16b or excessive carryover of liquid in the separated steam 16c by selectively decreasing or increasing it, respectively. For example, the steam and water mixture 16a at the top of the steam separator 28 is at a first pressure P.sub.1, the separated water 16b in the bypass channel 52 adjacent to the bypass inlets 58 is at a second pressure P.sub.2, and the pressure of the inlet water 16f just below the lower support plate 46 in the lower plenum 22 is at a third pressure P.sub.3. By predeterminedly selecting the level L of the water in the bypass channels 52, the second pressure P.sub.2 may be maintained greater than the third pressure P.sub.3 and less than the first pressure P.sub.1 to ensure travel of the separated water 16b from the steam separators 28 downwardly through the bypass channels 52 and into the lower plenum 22. Furthermore, in the embodiment of the invention without the core shroud 18 described above, the separated water 16b discharged through the bypass inlets 58 may also flow radially outwardly, or horizontally as shown in FIGS. 2 and 3, between adjacent baffles 50 and around the control rods 84 to the downcomer 20 (FIG. 1) and then downwardly into the lower plenum 22. The upwardly extending skirt 54 ensures that the water level L remains below the top of the skirt 54 and does not flow downwardly through the tops of the steam separators 28. In this way, the separated water 16b may be suitably channeled from the separators 28 and into the lower plenum 22 wherein any crud contained therein may be removed. More specifically, and as shown schematically in FIG. 1, conventional water clean-up means 76 are provided for separating crud from the separated water 16b flowing in the bypass channels 52. The separated water 16b flows through a plurality of bypass outlets 78 in the lower support 46, as shown in FIG. 3, and into the lower plenum 22. The crud removing means 76 include a conventional removal conduit 80 disposed in flow communication with the lower plenum 22 for removing a portion of the water therein, including the separated water 16b, which is channeled through a conventional filter for removing the crud therefrom. A conventional return conduit 82 is disposed in flow communication with the lower plenum 22 for returning the filtered water back into the lower plenum 22. Accordingly, the steam separators 28 may be used for withdrawing predetermined amounts of the separated water 16b for ensuring that any crud contained in the water remains in suspension therewith and may be channeled downwardly through the bypass channels 52 and into the lower plenum 22 so that the crud may be suitably removed. Buildup of the crud on the lower fuel bundles 26, as well as on the upper fuel bundles 32, may therefore be reduced or eliminated, thusly avoiding overheating the fuel bundles and adversely affecting the power distribution therefrom. Furthermore, the quality of the separated steam 16c channeled to the upper core 30 may be maintained at a predetermined value during the operating range of the reactor 10. For example, the upper core 30 may be optionally operated without conventional control rods, and using instead, as an example, posion curtains containing boron. The boron is essentially transparent to the hard neutron spectrum and will allow the upper core 30 to remain critical. However, at an off-design point wherein the density of the superheated steam 16d is below the density at the design point, the poison curtain will cause the upper core 30 to automatically become subcritical. The steam separators 28 limit the change in density range to ensure normal operation without inadvertent shutdown due to undesirable density changes. In the preferred embodiment illustrated in FIG. 1, a plurality of control rods 84 are disposed in the respective bypass channels 52 (see also FIGS. 2 and 3) for controlling reactivity of both the lower core 24 and the upper core 30. Since the fuel bundles 26, 32 and the baffles 50 are hexagonal in this exemplary embodiment, the control rods 84 are preferably Y-shaped in transverse section for vertical translation therebetween. In the exemplary embodiment illustrated in FIG. 1, a conventional control rod guide tube 86 extends downwardly from the lower core 24 and into the lower plenum 22, and a conventional control rod drive shaft 88 extends downwardly from the control rod 84 through the guide tube 86 and through the lower head of the vessel 12 into a conventional control rod drive 90. The several control rod drives 90 are conventionally effective for raising the shafts 88 upwardly for inserting the control rods into the respective cores for decreasing reactivity, and lowering the shafts 88 for withdrawing the control rods 84 from the respective cores for increasing reactivity. Illustrated in more particularity in FIG. 2, each of the control rods 84 preferably includes a first or lower portion 84a disposed parallel to and longitudinally coextensively with the lower fuel bundles 26, i.e. they have about the same longitudinal length L.sub.1 when inserted into the lower core 24. The lower portion 84a includes a conventional nuclear poison which reduces reactivity of the lower fuel bundles 26 when fully inserted into the lower core 24. Each of the control rods 84 also includes a second, or middle portion 84b extending vertically upwardly from the first portion 84 and being integral therewith, which is disposed parallel to the steam separators 28 and longitudinally coextensively therewith, i.e. having the same longitudinal length L.sub.3 when fully inserted. The middle portion 84b is inert, i.e. contains no nuclear poison, and is substantially transparent to neutrons. The middle portion 84b, along with the lower portion 84a, are also disposed in the bypass channel 52 when fully inserted. Each of the control rods 84 further includes a third, or top portion 84c extending vertically upwardly from the middle portion 84b and being integral therewith, and is disposed parallel to the upper fuel bundles 32 and extends longitudinally coextensively therewith, i.e. having about the same longitudinal length L.sub.2 when fully inserted into the upper core 30. The top portion 84c also contains a conventional nuclear poison for reducing reactivity in the upper fuel bundles 32 when fully inserted. As shown in FIG. 1, the exemplary centermost one of the control rods 84 is shown in its fully inserted position with the lower portion 84a being fully inserted in the lower core 24, and the upper portion 84c fully inserted in the upper core 30. An exemplary second one of the control rods 84 is shown in its fully withdrawn position with the top portion 84c being fully withdrawn from the upper core 30 and into the bypass channel 52 between the separators 28 and between the lower and upper cores 24, 30, and the lower portion 84a being withdrawn below the lower core 24 and into the lower plenum 22. The middle portion 84b is withdrawn from between the separators 28 and into the lower core 24. In this way maximum reactivity of the lower core 24 and the upper core 30 may occur. Alternatively, the reactor 10 could be reconfigured so that the control rods 84 are withdrawn from the respective cores 24 and 30 upwardly therefrom. In either embodiment, the control rod drives 90 are effective for selectively moving the control rods 84 for withdrawing to between the lower and upper cores 24 and 30 either the lower portion 84a or the upper portion 84c. In the preferred embodiment illustrated in FIG. 1, the control rod drives 90 are effective for moving downwardly the control rods 84 for withdrawing the upper portions 84c to the space between the lower core 24 and the upper core 30, and withdrawing the lower portion 84a into the lower plenum 22 below the lower core 24. And, the baffles 50 may additionally be used to guide the translation of the control rods 84 between adjacent ones of the respective fuel bundles 26, 32, and the steam separators 28. The bi-level reactor 10 including the steam separators 28 in accordance with the present invention, therefore, allows for more complete axial burnup of the nuclear fuel in the lower core 24 and the upper core 30 by taking advantage of the axially varying neutron density and spectrum and the axial reshuffling of the fuel bundles between the upper core 30 and the lower core 24. The steam separators 28 built into the lower and upper fuel bundles 26 and 32 allows for the recirculation of the separated water 16b to the lower plenum 22 for removing crud therefrom by the crud removing means 76. The steam separators 28 also reduce the variation in quality of the separated steam 16c thus resulting in higher quality inlet steam to the upper core 30 operating as a steam cooled reactor. By positioning the steam separators 28 between the lower core 24 and the upper core 30, the control rods 84 can be withdrawn axially, either upwardly from the lower core 24 to the region between the lower core 24 and the upper core 30 containing the steam separators 28, or, alternatively, downwardly from the upper core 30 into that region. While there have been described herein what are considered to be preferred embodiments of the present invention, other modifications of the invention shall be apparent to those skilled in the art from the teachings herein, and it is, therefore, desired to be secured in the appended claims all such modifications as fall within the true spirit and scope of the invention. Accordingly, what is desired to be secured by Letters Patent of the United States is the invention as defined and differentiated in the following claims: |
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abstract | Charged particles that are in transit through a deflection system when the beam is repositioned do not received the correct deflection force and are misdirected. By independently applying signals to the multiple stages of a deflection system, the number of misdirected particles during a pixel change is reduced. |
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046366458 | abstract | A closure system for a storage cask containing spent nuclear fuel has redundant mechanical seals to temporarily close the cask during development, testing, and refinement thereof, when it is occasionally necessary to open the cask, and redundant welded seals to supplement the mechanical seals when development is completed and the cask is permanently sealed during long-term storage. The closure system includes a primary cover which cooperates with an O-ring to provide a mechanical seal and which includes a canopy element that can be welded to the cask base element if a permanent seal is desired. The closure system also includes a secondary cover which cooperates with an additional O-ring to provide an additional mechanical seal and which is configured to receive an additional canopy element in the event that an additional welded seal is desired. The closure system permits the cask to be developed, tested, and refined using the same closure system that will be installed during actual use. |
abstract | The present invention provides a vapor-cell system comprising: a vapor-cell region configured for vapor-cell optical paths; a first electrode disposed in contact with the vapor-cell region; a second electrode electrically isolated from the first electrode; and an ion conductor interposed between the first electrode and the second electrode. The first electrode, the ion conductor, and the second electrode collectively form a bidirectional solid-state electrochemical charge-depletion capacitor. The ion conductor is ionically conductive for mobile ions, such as Rb+, Cs+, Na+, K+, or Sr2+. The first electrode is permeable to the mobile ions and/or neutral atoms formed from the mobile ions. The system can be electrically controlled to quickly pump mobile ions into or out of the vapor-cell region. The system may further contain an atom chip, and the vapor-cell optical paths may be configured to trap a population of cold atoms. Methods of operating these vapor-cell systems are also disclosed. |
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044328940 | description | DETAILED DESCRIPTION OF THE INVENTION The present invention is characterized by mixing an adsorbent with a radioactive liquid waste containing a surface active agent and thereafter subjecting the waste to concentration and drying. A satisfactory powdery product having a low moisture content is obtained by mixing powdery or particulate activated carbon with a detergent-containing radioactive liquid waste and then subjecting the mixture to the action of a thin film evaporator; it has been experimentally confirmed that properties, and particularly the moisture content of the product depend on the mixing ratio of activated carbon to the chemical oxygen demand (COD) of the radioactive waste. The present invention is based on such finding. The experimental results are given below. Activated carbon is mixed with a radioactive waste of 10 wt. % detergent concentration, and then the mixture is fed to a thin film evaporator, 2 m.sup.2 in heating surface area at a flow rate of 160 kg/hr to be evaporated and dried at 170.degree. C. FIG. 1 shows the moisture content of the powder obtained by the above treatment varied with the activated carbon mixing ratio. The latter means [concentration (wt. %) of activated carbon in the waste]/[COD concentration (wt. %) of the waste]. It is easily seen that the moisture content of the resulting powder lowers with increasing activated carbon mixing ratio, and this result attests to the effectiveness of activated carbon incorporation, because the surface active agent and other ingredients of the detergent contained are adsorbed by the activated carbon having a large surface area, and their undesirable effect are removed. A similar effect was obtained by the use of other adsorbents having large surface area, such as molecular sieve, silica gel and alumina. FIG. 1 indicates that the mixing ratio is preferably not less than 1 because the moisture content of the resulting powder increases rapidly when the mixing ratio is less than 1. It is more preferable that the mixing ratio is 2-4, as a range for a remarkable decrease in the moisture content below 15 wt. % by addition of activated carbon. Furthermore it has been found that the volume of the resultant powder or pellets can be made as small as possible at the mixing ratio of about 3. It has been experimentally confirmed that the powdery product resulting at the mixing ratio can easily be pelletized into pellets of high strength suitable enough for handling and storing. According to the experiments, pellets shaped without the activated carbon addition have a breaking strength of about 60 kg/piece, whereas pellets with the activated carbon at the mixing ratio of 3 have an improved strength of about 100 kg/piece. FIG. 2 shows a flow diagram of an apparatus for the treatment of radioactive liquid waste suitable for carrying out the present invention. Waste tank 16 containing radioactive liquid waste 10 with a surface active agent connected to an ejector 22 through piping 19 provided with a pump 20. Tank 24 containing powdery or granular activated carbon 26 is connected to the ejector 22 through a piping 27 provided with a device 28 for adjusting the amount of the activated carbon. Pipe 29 connects the ejector 22 to a mixing tank 30 provided with a stirrer 32. Pipe 35 provided with a feed pump 34 and a flow meter 36 connects the mixing tank 30 to a thin film evaporator 38. Rotary shaft 46 is provided inside vessel 40 of the thin film evaporator 38, and wiping blades 48 are rotatably provided at the rotary shaft 46. A jacket 44 is provided at the outside of vessel 40. Steam feed pipe 42 is connected to the jacket 44. Numeral 50 is a pelletizer and 54 is a drum. Surface active agent-containing radioactive liquid waste 10 (e.g., laundry drainage) generated in a boiling water-type atomic power plant is concentrated in a reverse osmosis apparatus 12, and then further concentrated in a concentrator 14 to a predetermined concentration, e.g., 10 wt. %, and then fed to the waste tank 16. The COD concentration of the waste 10 is measured by a COD meter 18 provided inside the waste tank 16. A commercially available automatic COD measuring device can be used as the COD meter. Then, the waste 10 is fed to the ejector 22 by driving a transfer pump 20. By the action of ejector 22 granular activated carbon 26, about 250 microns in particle size and 800 m.sup.2 /gr in specific surface area, added to the waste 10 from the tank 24. The device 28 for adjusting the amount of activated carbon is adjusted so that the amount of activated carbon 26 corresponding to the measured COD concentration can be fed to the ejector 22. The waste 10 containing activated carbon 26 is transferred to the mixing tank 30 and mixed further by the action of the stirrer 32. Thereafter, by driving the feed pump 34, the waste is fed into the vessel 40 of the thin film evaporator 38 through the flow meter 36. The surface of the container wall is heated by the steam fed into the jacket 44 through the steam feed pipe 42. As the rotary shaft rotates, wiping blades 48 travel in the circumferential direction of the vessel 40 along the inner wall; whereas the waste 10 is pressed against the inner wall to form a thin film thereon by the centrifugal force of the blades 48 and then flows down along the inner wall while heated. The waste 10 is concentrated, dried and converted to powder by the action of the blades 40. It can easily be pulverized because the surface active agent contained in the waste is adsorbed on activated carbon 26. The activated carbon 26 is also powdered by the wiping blades 48. The resulting powdery product is taken out of the vessel 40 at its bottom. This product containing activated carbon is shaped into pellets 52 by a pelletizer 50 and then packed in drums 54. Then, asphalt or a plastic is poured into each drum 54 to solidify the powder. Steam from the evaporator 38 is led to a condenser 56, cooled by cooling water 58, and recovered in a condensate water tank 60 as condensate 62. Because of the porosity of the activated carbon, the powder prepared from the waste 10 enters the pores when pelletized to form pellets of high strength. Description will be made below of a case where one cubic meter of the radioactive waste containing about 10 wt. % of a detergent and a COD concentration of 0.7 wt. % is treated in the apparatus of FIG. 2. If the activated carbon mixing ratio of 3 is selected for safety on the basis of the experimental results shown in FIG. 1, the concentration of the activated carbon 26 to be added will be 2.1 wt. %, i.e. three times the COD concentration of the radioactive waste (0.7 wt. %). 21 kg of granular activated carbon 26 is added to the waste 10 from the tank 24 and fully mixed in the mixing tank 30. Then, the mixture is introduced into the evaporator 38 at a rate of 200 kg/hr. A powdery product is obtained at a rate of about 20 kg/hr from the evaporator 38 at 170.degree. C., and the waste 10 is converted to powder. The moisture content of the powdery product was about 10 wt. %. The COD concentration includes such factors as type of surface active agent and the degree of deterioration thereof. A proper amount of activated carbon can be added because the amount of granular activated carbon to be added to the radioactive waste is adjusted on the basis of the COD concentration of the waste containing surface active agent. The amount of activated carbon to be added varies with the type of surface active agent and the degree of deterioration thereof. Accordingly, if the amount of activated carbon is selected solely on the basis of the concentration of detergent or surface active agent, too a small or too a large amount is inevitably added and the proper amount cannot be added. Too much activated carbon increases the amount of the resultant radioactive waste, whereas in the present invention the amount of activated carbon added is so small that the increase in an amount of the resultant waste is very small. By converting the radioactive liquid waste containing the surface active agent to the form of powder its volume can be decreased remarkably. Besides, the resulting powder has a low moisture content and can easily be pelletized, so its volume can be made even smaller. The foregoing example of the invention is directed to the treatment of radioactive liquid waste containing detergents alone. However, in the boiling water-type atomic power plants are generated not only the detergent-containing radioactive waste but also slurry-type radioactive waste containing used powdery ion exchange resin, sodium sulphate-containing radioactive waste resulting from regeneration of used ion exchange resin, etc. Thus, the present invention is applicable to (1) a mixture of radioactive liquid waste containing surface active agents and radioactive liquid waste containing sodium sulphate and (2) the mixture of (1) mixed further with slurry-type radioactive waste containing used powdery ion exchange resin. Activated carbon is added to the mixture (1) or (2) and then the resulting mixture is converted into powder in the thin film evaporator of FIG. 2. Then, the moisture content of the powder is measured. The results are shown in the following Table. The concentration of surface active agents in the detergent is 14 wt. % in each case. TABLE ______________________________________ Moisture content Percentage of ingredients of resul- in the waste (wt. %) tant Exp. Powdery ion powder No. Detergent Na.sub.2 SO.sub.4 exchange resin (wt. %) ______________________________________ 1 10 90 0 7.4 2 25 75 0 11.0 3 50 50 0 12.7 4 75 25 0 13.5 5 100 0 0 14.0 6 34 33 33 3.5 7 66 17 17 5.2 8 80 10 10 7.4 9 90 5 5 10.0 ______________________________________ It is seen from the Table that all the powder produced by treating radioactive liquid wastes of varied compositions have moisture contents of not more than 14 wt. %, and that by the addition of activated carbon the mixed waste can be converted into powder easily with satisfactory result. It must be because part of the surface active agents which are main ingredient of the detergent, that is, ionic surface active agents, are retained on the surface of the powdered ion exchange resin owing to ion exchange, etc. that the resulting powder has a lower moisture content when the mixed waste to be treated contains a powdery ion exchange resin. FIG. 3 shows comparison of the case where activated carbon is added to the mixed waste (1) with the case where no activated carbon is added to the same waste in respect to the moisture content of the powder produced; and FIG. 4 shows comparison of the case where activated carbon is added to the mixed waste (2) with the case where no activated carbon is added to the same waste in respect to the moisture content. In FIGS. 3 and 4, curves I and III are characteristic curves obtained when activated carbon is added according to the present invention; whereas curves II and IV are characteristic curves of the prior art in which no activated carbon is added. In the prior art cases, as shown by curves II and IV, the mixed waste can be converted into powder when its detergent content is low because the detergent ingredients are, to some extent, retained by sodium sulphate and the powdery ion exchange resin; but in case the detergent content exceeds 5 wt. %, the moisture content of the product obtainable by the treatment of the mixed waste increases so rapidly, and the powder formation cannot be attained. On the other hand, in the present invention, a satisfactory powdery product not exceeding 14 wt. % in moisture content can be obtained in every case, inclusive of the case where the radioactive waste to be treated contains a detergent alone. If the moisture content of the resulting powder exceeds 15 wt. %, the amount of moisture on the surface of powder particles increases so as to make pelletizing difficult. According to the present invention, detergent-containing radioactive liquid wastes of various compositions generated in atomic power plants, etc. can easily be converted into powder, and the resulting powder has a low moisture content and is therefore suitable for pelletizing, and thus the amount of radioactive waste to be stored can be minimized. Activated carbon serves as a buffer between the wall surface and the wiping blades and lessens abrasion of the blades of the thin film evaporator. Other adsorbents as described above have also an effect of lessening the abrasion. As described before, molecular sieve, silica gel, alumina, etc. may be used in place of activated carbon and it is desirable that it should have a large surface area. If the surface area is small, the amount thereof to be added will increase. The addition of the adsorbent is made in accordance with the COD concentration of the radioactive waste in the foregoing examples, but by fixing the maximum content (or maximum COD concentration) of the detergent contained therein to a predetermined value, a necessary amount of adsorbent can be added on the basis of the fixed content or concentration. In the latter case, operation can be made very simple but there appears such a disadvantage that the amount of adsorbent to be added will increase. A thin film evaporator is used as the apparatus for evaporation and drying in the foregoing embodiments, but the invention can be effectively carried out by pulverization by other means for concentration and drying, such as fluidized-bed concentration and drying. The resulting particle size and other properties inevitably vary with the drying system as used. It will be needless to say that the present invention is generally applicable to liquid wastes of chemical plants, though the waste to be treated has been limited to radioactive waste originating from atomic power plants in the foregoing embodiments. In a word, radioactive liquid wastes containing surface active agent can be converted into powder of low moisture content, resulting in remarkable decrease in its volume according to the present invention. |
abstract | A control system for the operation of a centrifugal pump which may be used for production of gas and/or oil from a well. The control system includes vector feedback model to derive values of torque and speed from signals indicative of instantaneous current and voltage drawn by the pump motor, a pump model which derives values of the fluid flow rate and the head pressure for the pump from torque and speed inputs, a pumping system model that derives from the estimated values of the pump operating parameters an estimated value of a pumping system parameter and controllers responsive to the estimated values of the pumping system parameters to control the pump to maintain fluid level at the pump input near an optimum level. |
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048620050 | abstract | An apparatus for detecting radioactive contamination in hand-held objects, such as tools used to service a nuclear power facility, is disclosed herein. The apparatus generally comprises a radiation detector assembly having a gas-flow proportional detector, and a platform assembly formed from a perforated sheet of rigid material disposed over the topside of the detector for both supporting the hand-holdable objects and uniformly spacing the object from the detector. The radiation detector assembly is contained within a shielding cabinet having an access opening that is offset out of alignment with the top side of the detector for allowing an operator to deposit and withdraw an object onto and off of the platform. The walls of the shielding cabinet include pocket-like mounting assemblies for releasably holding one or more sheets of lead shielding material so that the amount of background radiation-reducing shielding may be advantageously adjusted. The shielding cabinet further includes a pair of cabinet doors on opposite sides of the cabinets which swing down into a horizontal position which is substantially in alignment with the top surface of the detector. The extra access openings provided in the cabinet, along with the shelf-like support afforded by the swung-down cabinet doors allows objects that are longer than the width of the cabinet to be easily drawn over the top surface of the detector and scanned thereby. Finally, the apparatus includes a space gas-flow proportional radiation detector that undergoes a constant purging with counting gas so that it may immediately replace the primary detector in the event of a malfunction, thereby minimizing downtime. |
047818840 | abstract | A debris catching strainer grid for capturing and retaining deleterious debris carried by reactor coolant before it enters the active region of a fuel assembly and creates fuel rod cladding damage has a plurality of fuel end cap compartments defined by pairs of first and second intersecting and slottedly interlocked grid-forming strips attached to a perimeter member and to each other. The end cap compartments defined by strips including vertical rows of integral leaves on opposite sides of the strip or by pairs of adjacent integral leaves intermediate their intersections. In the latter case, each leaf of the pair of leaves is the mirror image of the other leaf of the pair with an asymmetric shape with the greatest distance of projection out of the plane of the strip remote from the midpoint of the strips between their intersections. |
039752330 | claims | 1. A safety rod for a neutronic reactor comprising an elongated hollow tube, a first region approximately nine feet in length comprising a plurality of spaced layers of "masonite" disposed within the tube adjacent to one end thereof, and a layer of soft steel disposed within the tube between each pair of spaced layers of masonite, a second region adjacent to the first region comprising a sleeve disposed within the tube having a smaller outer diameter than the inner diameter of the tube forming a space between the sleeve and the tube, powdered boron disposed within said space, and a plurality of discs of polyethylene disposed within the sleeve, and a third region approximately eleven and one-half feet in length adjacent to the second region comprising a lead plug within the tube and discs of polyethylene adjacent to both sides of the plug. 2. A neutronic reactor comprising, in combination an active portion containing material fissionable by neutrons of thermal energy provided with a channel therein adapted to receive a neutron absorbing rod, a radiation shield surrounding the active portion of the reactor having an aperture aligned with said channel, said shield having an iron thermal shield approximately 10 inches thick surrounded by a laminated biological shield approximately 491/3 inches thick, said biological shield having alternate layers of "masonite" approximately 41/2 inches thick and iron approximately 33/4 inches thick, a neutron absorbing rod slidably disposed within the channel and aperture in the shield, said rod having an inner end portion approximately 9 feet in length containing layers of "masonite" approximately 131/2 inches long and layers of soft steel approximately 33/4 inches long, and said rod having an outer end portion approximately 111/2 feet long containing a lead cylinder approximately 6 inches in length and two abutting polyethylene discs, the disc between the lead cylinder and the outer end of the rod being approximately 8 inches in length, and the other disc being approximately 114 inches in length. 3. A neutronic reactor having an active portion provided with a channel therein adapted to receive a neutron absorbing rod, a radiation shield surrounding the active portion of the reactor and having an aperture aligned with said channel, and a neutronic absorbing rod slidably disposed within said channel characterized by the construction wherein the rod comprises an inner end portion at least equal in length to the thickness of the shield and normally positioned within said aperture, said inner end portion having a neutron capture cross section and a gamma ray absorption coefficient at least equal to those of the surrounding shield, a central portion having an axial portion containing material having a neutron slowing power of at least 0.1 and a peripheral portion containing material having a neutron capture cross section of at least 100 barns, and an outer end portion at least equal in length to the thickness of the shield and having a gamma ray absorption coefficient at least as great as that of the shield. 4. A neutronic reactor comprising the elements of claim 3 wherein the neutron absorbing material in the central portion of the neutron absorbing rod comprises boron, and the neutron slowing material comprises polyethylene. |
048428108 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The invention relates to a nuclear power plant and, more particularly, to a plant with a cylindrical pre-stressed concrete pressure vessel containing an eccentrically arranged reactor. The pressure vessel cavity is clad with a liner. The reactor is a helium cooled high temperature reactor with a pile of spherical fuel elements arranged in its core. A graphite reflector surrounds the pellet pile on all sides and a plurality of absorber rods are insertable into the lateral portion of the graphite reflector. A cold gas collector chamber is located above the hight temperature reactor; at least two parallel heat exchangers are arranged in the cavity and the same number of circulating blowers follow the heat exchangers. The heat exchangers are installed adjacent to the high temperature reactor in the cavity and upwardly offset from the reactor. The heat exchangers are intended for operational heat removal and discharge of decay heat. A liner cooling system may also be used for the removal of decay heat. 2. Description of the Related Technology West German Application DE P No. 36 21 516.3 corresponding to U.S. application Ser. No. 066,464 incorporated by reference herein shows a nuclear power plant with a power rating of 50-300 MWth. Heat is removed by several intermediate circulating loops and is used to supply a remote heating system. Each intermediate circulating loop comprises an intermediate heat exchanger and a circulating pump. An auxiliary loop is installed parallel to each intermediate circulating loop. The auxiliary loop is shut off in normal operation and contains a recooling system. The plant exhibits an elevated reservoir filled with water and connected to a further heat sink. The auxiliary circulating loops and the recooling system serve to remove the decay heat in the case of an accident. West German Application DE P No. 35 34 423.7 shows a nuclear reactor plant with a small high temperature reactor. It proposes to connect the liner cooling system of the prestressed concrete pressure vessel housing the small high temperature reactor to an elevated reservoir filled with water under atmospheric pressure so that the liner cooling system may be operated by natural circulation. West German Application DE-OS No. 34 35 255 shows a nuclear power plant with a small high temperature reactor installed in a steel pressure vessel. The removal of the decay heat is effected by special decay heat exchangers connected on the cooling water side to an external, geographically elevated recooling heat exchanger. SUMMARY OF THE INVENTION It is an object of the invention to provide a nuclear power plant with an output capacity of 100-500 MWth and primarily for power generation, in which case the possibility of the release of radioactivity is excluded even in the case of extremely severe accidents. The plant with a cylindrical prestressed concrete pressure vessel defining a liner clad cavity. A high temperature reactor is located eccentrically in the cavity exhibiting a core with a pile of spherical fuel elements. Helium coolant flows downward through the core and a graphite reflector surrounds the pellet pile on all sides. A plurality of absorber rods are insertable into the pile and into the lateral graphite reflector. A cold gas collector chamber is located above the high temperature reactor. At least two parallel heat exchangers are located in the cavity with corresponding circulating blowers following the heat exchangers. The heat exchangers are installed adjacent to the high temperature reactor in the cavity and offset upwardly from the high temperature reactor. The heat exchangers are intended simultaneously for the operational removal of heat and removal of decay heat. A liner cooling system may be used for the removal of decay heat. The nuclear power plant should be economically manufactured and operated. The secondary side of each heat exchanger is connected to a water-stream loop for power generation. Advantageously, a cold gas chamber has a thermal capacity for hot gas temperatures and the liner cooling system and thermal insulation applied on the inside to the prestressed concrete pressure vessel may be arranged for removal of decay heat at an elevated temperature level. The liner cooling system may be connected to an elevated reservoir filled with water through an external cooling loop. The reservoir may further be connected to an additional heat sink by a recooling circulating loop. A device may be provided for supplying cooling water through the external loop to the liner cooling system. The proposed nuclear power plant provides a passively safe high temperature reactor utilizing natural convection in combination with a liner cooling system as installations for removal of decay heat. Evacuation and relocation of the population living in the vicinity of the plant may be avoided. An accident of this type may occur if the heat exchangers or the circulating blowers fail. Decay heat may be removed by the liner cooling system alone by natural convection while the reactor is under pressure. The removal of decay heat may be continued for several days by evaporation of the water supply contained in the connected elevated reservoir. Further, additional water may be fed into the elevated reservoir and thus into the liner cooling system. If, in the course of an accident, the cooling medium is also lost, decay heat is conveyed to the liner cooling system by conduction and radiation. The plant and pressure vessel are designed with sufficient integrity so that radioactivity remains safely contained in the prestressed concrete pressure vessel upon failure of the decay heat removal systems even in hypothetical accidents. An advantage of the invention is that there is no risk of loss of capital in the case of the nuclear power plant according to the invention. A further advantage resides in the use of a prestressed concrete pressure vessel in place of the steel pressure customarily provided in plants of this capacity. The nuclear power plant of the invention is of interest particularly for countries which are not capable of manufacturing large steel vessels or transporting them over long distances. Furthermore, numerous structural parts known from the THTR-300 may be used in such a nuclear power plant, so as to minimize expenditures for new developments. The circulating blowers may be vertically installed above their associate heat exchanger in a passage in the prestressed concrete pressure vessel. Each heat exchanger may advantageously be equipped with an external device provided for the introduction of feed water. The configuration and prestressing system of the prestressed concrete pressure vessel are such that the pressure vessel passage closures are not unduly stressed. Advantageously, the installation of thermal insulation is optimized for minimal heat transfer during power operation and optimum heat transfer during the removal of decay heat. The capacity of the nuclear power plant may be increased by multiplication or addition of heat exchanger/blower units alone, while maintaining the same basic configuration. Further, advantageous features of the invention will become apparent from the description below of an embodiment with reference to the schematic drawing . |
claims | 1. An imaging apparatus, comprising:a radiation source, including:a cathode configured to emit an electron beam;an anode, including:a plurality of radial slits; anda target layer disposed on a surface of the anode between the radial slits,wherein the target layer is configured to emit X-ray radiation in response to receiving the electron beam; anda drive unit configured to rotate the X-ray anode during scanning such that during a first time interval the electron beam passes through the plurality of radial slits and during a subsequent time interval the target layer receives the electron beam and emits the X-ray radiation; anda radiation detector configured to generate a first electrical signal during the first time interval and detect the emitted X-ray radiation and generate a subsequent electrical signal indicative thereof during the subsequent time interval. 2. The imaging apparatus of claim 1, further comprising:first circuitry configured to generate a first detector signal indicative of a first number of photons of the received X-ray radiation during the first time interval from the first electrical signal. 3. The imaging apparatus of claim 2, further comprising:second circuitry configured to sense a persistent electrical current in the first signal. 4. The imaging apparatus of claim 3, wherein the first circuitry is further configured to generate a subsequent detector signal indicative of a subsequent number of photons during the subsequent time interval from the subsequent electrical signal. 5. The imaging apparatus of claim 4, wherein the second circuitry is further configured to correct, based on the persistent electrical current, the subsequent detector signal. 6. The imaging apparatus of claim 5, further comprising:third circuitry which energy-discriminates and bins the corrected subsequent detector signal; anda reconstructor that reconstructs the energy-binned signal to produce an image. 7. The imaging apparatus of claim 1, wherein a slit of the plurality of radial slits includes a width, which is equal to FS+(R×Ω×T), wherein the parameter FS is a focal spot size on the anode, the parameter R is a radius of a focal track on the anode, the parameter Ω is an angular speed of the anode, and the parameter T is a minimum time required to sense the persistent electrical current. 8. The imaging apparatus of claim 7, wherein the parameter R is in a range from 5 centimeters to 35 centimeters, the parameter Ω is in a range from 2π×50 Hertz to 2π×400 Hertz, and the parameter T is in a range from 0.1 micro-seconds to 100 micro-seconds. 9. The imaging apparatus of claim 7, wherein the width is in a range from 0.5 millimeter to 3 millimeters. 10. The imaging apparatus of claim 5, wherein the second circuitry comprises:sample and hold circuitry configured to receive the first detector signal and generate a persistent current compensation signal based on the persistent electrical current, and wherein the second circuitry corrects the subsequent electrical signal with the persistent current compensation signal. 11. The imaging apparatus of claim 10, wherein the sample and hold circuitry comprises:a switch configured to enable dynamic adjustment of the subsequent electrical signal with the persistent current compensation signal. 12. The imaging apparatus of claim 11, wherein the switch is controlled to switch synchronously with the first and subsequent time intervals. 13. The imaging apparatus of claim 11, wherein the switch is controlled to switch asynchronously with the subsequent time interval, where a reference signal corrects for a variation in x-ray flux. 14. The imaging apparatus of claim 11, wherein the subsequent time interval is a measurement interval, and the switch is controlled to switch synchronously with the first time interval. 15. A method, comprising:rotating a slotted anode of a radiation source, wherein the slotted anode includes a plurality of radial slits with a target layer disposed on a surface of the anode in between the plurality of radial slits;emitting, concurrently with rotating the slotted anode, an electron beam with a cathode of the radiation source;wherein the radiation source emits an X-ray beam during periods in which the electron beam is received by the target layer and does not emit the X-ray beam during periods in which the electron beam passes through the plurality of radial slits;detecting, with a radiation detector, the emitted X-ray radiation; andgenerating a first electrical signal indicative thereof during the periods in which the electron beam is received by the target layer and generating a second electrical signal during the periods in which the electron beam passes through the plurality of radial slits. 16. The method of claim 15, further comprising:sensing a persistent electrical current of the radiation detector from the second electrical signal. 17. The method of claim 16, further comprising:correcting the first electrical signal with the sensed persistent electrical current. 18. A non-transitory computer readable medium encoded with computer executable instruction which when executed by a processor cause the processor to:rotate a slotted anode of a radiation source, wherein the slotted anode includes a plurality of radial slits with a target layer disposed on a surface of the anode in between the plurality of radial slits;emit, concurrently with rotating the slotted anode, an electron beam with a cathode of the radiation source;wherein the radiation source emits an X-ray beam during periods in which the electron beam is received by the target layer and does not emit the X-ray beam during periods in which the electron beam passes through the plurality of radial slits;detect, with a radiation detector, the emitted X-ray radiation; andgenerate a first electrical signal indicative thereof during the periods in which the electron beam is received by the target layer and generate a second electrical signal during the periods in which the electron beam passes through the plurality of radial slits. 19. The non-transitory computer readable medium of claim 18, wherein executing the computer executable instructions further causes the processor to:rotate the slotted anode synchronously with data acquisition intervals. 20. The non-transitory computer readable medium of claim 18, wherein executing the computer executable instructions further causes the processor to:rotate the slotted anode asynchronously with data acquisition intervals. |
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048511838 | description | DETAILED DESCRIPTION OF THE INVENTION Referring now to FIG. 1, the reactor core is disposed within a comparatively low-cost pressure vessel (10) at the bottom of a vertical shaft (20). The shaft is of sufficient diameter to receive the horizontal cross section of the reactor pressure vessel and may possibly but not exclusively have a depth of 600-1500 feet. That depth which constitutes a safe distance for isolating nuclear contaminants from the atmosphere depends on details of the local lithology and stratigraphy. A possible depth cited in the technical literature for underground siting of a conventional nuclear power plant is 340 feet (Bowman, Watling, and McCauley, op. cit.). To be safe against the worst conceivable reactor accident, namely the nearby explosion of a nuclear weapon, the reactor must be situated at a depth of at least 800 feet. The shaft (20) is lined with a borehole liner (30) of impermeable material which in a preferred embodiment of the invention comprises a concrete-encased, thermally insulated steel pipe. Heat transfer means (40) remove heat from the reactor core and transport it to heat utilization means (50) located at or near the surface of the earth. The heat utilization means (50) may typically comprise a boiler, turbine generator, and cooling tower. In one possible embodiment of the invention, as shown in FIG. 1, the means (50) for exchanging and utilizing heat from the reactor occupy a central generating facility on or near the surface of the earth, where they can readily be serviced like those of any conventional coal- or gas-fired power plant, and where they do not require any extra and costly pressure containment vessels and safety cooling system. The heat exchange and utilization means (50) at the surface of the earth are surrounded by a plurality of shafts (20), (60), (70). At any one time during the operative lifetime of the invention, one or more operative reactors at the bottoms of their respective shafts are thermally connected to the centrally located heat exchange and utilization means (50) by underground, vertical heat pipes which are thermally connected to the central heat exchange means (50). As an individual reactor completes its operative lifetime, it is cut off from the heat exchange means and sealed in situ within the lower portion of the casing by activating valves and underground mechanical closures (80) and possibly explosive closures (90). Thus the old reactor is abandoned in place in a deactivated hole (60). New reactors can be installed in previously unused shafts (70). Prior to becoming operative, new reactors are thermally connected to the central heat exchange means and electrically heated to liquefy the solidified working fluid and the liquefiable neutron-reflecting material. Referring now to FIG. 2, the reactor is of the type known in the art as a "self-regulating, heat-pipe controlled, reflector-critical, compact, fast reactor." The reactor comprises a core (100) which in a preferred embodiment is conical, and nested, inert-gas buffered heat pipes (110). The heat pipes (110) are arranged preferably in primary, secondary, and higher-order arrays. The primary heat pipes, which extend into the core of the reactor and remove heat directly therefrom, are conical in the preferred embodiment of the invention. The evaporator sections (120) of the primary heat pipes are received within the reactor core (100). The condensor sections (130) of the primary heat pipes are received within the evaporator sections (140) of the secondary heat pipes. Similarly, the condensor sections of the secondary heatpipes may be received coaxially within the evaporator sections of the tertiary heat pipes, and so forth. The heat pipes emerging from the reactor core pass through a reflector region and extend vertically in the space above the reactor. The reflector region comprises a liquid-reflector reservoir (145) and a neutron-reflecting mantle (150). A liquefied neutron-reflecting material is transferred to the liquid-reflector reservoir from a storage reservoir (155). The design and use of a reactor of this kind is described in V. Hampel, U.S. Defensive Publication No. T101,204, "Compact Fast Nuclear Reactor Using Heat Pipes." The pressure vessel (10) enclosing the core is composed of heat-resistant material chosen to maintain its structural strength at the operating temperature of the reactor core, which possibly but not exclusively may lie in the range between 1400 and 2500 Kelvin degrees. The operating pressure lies in the range of vapor pressures of suitable working fluids employed in heat pipes in the range of operating temperatures. A partial list of possible working fluids includes lithium fluoride, lithium, and beryllium difluoride. Over the temperature range from 1400 to 2500 Kelvin degrees, the vapor pressure of lithium fluoride ranges from 0.01 to 15 bars. Over the same temperature range, the vapor pressure of lithium ranges from 0.1 to 50 bars. Over the same temperature range, the vapor pressure of beryllium difluoride ranges from 1 to 10,000 bars. The pressure vessel (10) is continuous with a casing (160) which extends vertically upward within the borehole liner between the reactor core and the surface of the earth. The pressure vessel (10) is enclosed within and thermally insulated from the borehole liner (30). A thermally insulating layer (170) is disposed within the annular space (180) defined between the outer surface of the pressure vessel and the inner surface of the borehole liner. In a preferred embodiment of the invention, the insulating layer comprises multiple layers of reflective foil in an evacuated space. In an alternative embodiment, the annular space (180) may additionally contain a liquid coolant for circulation as a carrier of low-grade heat in a cogeneration loop. The annular space (180) additionally contains temperature sensors and chemical sensors to assure that the contents of the pressure vessel and casing are thermally insulated from the ground and that there is no exchange of material between the pressure vessel or casing and the ground. A series of at least two thermally connected heat pipe arrays (190) (i.e., the primary and secondary arrays) extends within the casing (160) from the reactor core to heat utilization means (50) situated at or near the surface of the earth. Additional stages of heat-pipe arrays may be interposed within the casing between the secondary heat-pipe array and the heat-exchange means. Each higher stage is added by coaxially receiving the condenser section of a lower-stage heat pipe within the evaporator section of the corresponding next-higher-stage heat pipe. Referring now to FIG. 3, heat may additionally be transferred between vertical heat-pipe stages by the use of a thermal coupling manifold (200). The manifold (200) is a thermally insulated enclosure (210) filled with a heat-conductive fluid (220). The condensor ends (230) of the lower-stage heat pipes enter from the bottom of the manifold and terminate within the heat-conductive fluid. The evaporator ends (240) of the higher-stage heat pipes terminate within the heat-conductive fluid and exit through the top of the manifold. This arrangement decouples every individual heat pipe of a higher-stage array from any specific heat pipe in the lower-stage array. This offers distinct advantages in the case of failure of a heat pipe. In that case, the remaining heat pipes share the load previously carried by the failed heat pipe. In a preferred embodiment of the invention, semi-helical baffles (250) within the heat pipes deflect the evaporating gas as it condenses and drive the condensate towards the evaporator end of the heat pipes. This pumping action enhances the passive gravitational return of the heat-pipe fluid, reducing the likelihood of burnout over a range of power levels, and thus providing substantially fail-safe operation. The condensor section of the highest-stage heat pipe array is thermally connected to the heat-utilization means. Referring now to FIG. 4, the heat pipes in one or more arrays may contain, in addition to the working fluid, quantities of inert buffer gas to enhance the operation of the passive self-controlling mechanism and to provide remote, fast-acting, active reactor power control. The action of the buffer gas which may enhance passive self-control is described by Hampel (U.S. Defensive Publication No. T101,204, "Compact Fast Nuclear Reactor Using Heat Pipes"), and has been used in radioisotope space-power heat sources to regulate the temperature of the thermionic diodes over time. A preferred embodiment of the invention additionally incorporates active means, essentially as described by Hampel (U.S. Defensive Publication No. T101,204, "Compact Fast Nuclear Reactor Using Heat Pipes"), to control the reactivity of the reactor core by adjusting the pressure of buffer gas in one or more heat pipe arrays. Accordingly, a preferred embodiment of the invention additionally comprises high- and low-pressure gas lines (260) extending vertically within the casing (160) between the vertical heat pipes to means located at suitable depths below the surface of the earth for supplying, storing, and controlling the inert buffer gas. Each pair of high-pressure and low-pressure gas lines terminates at its lower end in a three-way valve (270) communicating with a gas-flow inlet (280) through the wall of a secondary or higher-stage heat pipe (290) near the condensor section of said heat pipe. The setting of the valve can be electro-pneumatically adjusted to either open the inlet from the high-pressure gas line to the heat pipe, effectively raising the operating temperature of the heat pipe, or to open the low-pressure inlet, permitting the venting of buffer gas and effectively lowering the operating temperature, or to close off both gas lines from the heat pipe. Sensors within the pressure vessel, casing, annular space, heat pipes, and manifold transmit information about temperature, pressure, chemical composition, and other operating parameters to a control center at or near the surface of the earth. Referring now to FIG. 5, at various depths along the shaft there are disposed closure means (90) and (80) to sealingly close off the casing and all heat pipes and electro-pneumatic control lines and sensors contained therein. The positions of the closure means are chosen to be optimally effective in isolating from the atmosphere such gaseous and particulate contaminants as might issue from the reactor system in the case of malfunction, and as might be expected to issue during the cooling-off period of a reactor that has been permanently shut down. In a preferred embodiment of the invention, the closure means include first-acting high-explosive-actuated pipe closures (90), and later-acting mechanically or pneumatically driven butterfly valves, sliding gates, and/or miter valves (80). The mechanically or pneumatically driven closure means (80) are disposed between the explosive closures (90) and the surface of the earth. The foregoing description has been presented for purposes of illustration and description. It is not intended to be exhaustive or to limit the invention to the precise forms disclosed, and obviously many modifications and variations are possible in light of the above teaching. For example, the reactor may employ a cylindrical core and cylindrical heat pipes in place of the conical elements described herein. The embodiment was chosen and described in order to best explain the principles of the invention and its practical application to thereby enable others skilled in the art to best utilize the invention in various embodiments and with various modifications as are suited to the particular use contemplated. It is intended that the scope of the invention be defined by the claims appended hereto. |
description | This application is a continuation-in-part of application Ser. No. 09/346,902 filted Jul. 2,1999 now U.S. Pat. No. 6,807,314. The present invention relates to a method for precision magnification calibration of image magnifying devices, such as optical microscopes, confocal scanning microscopes, transmission electron microscopes, scanning electron microscopes, tunneling microscopes and atomic force microscopes. Technical advances in many scientific fields are placing demands for increasing accuracy of feature measurements made using image-magnifying instruments [micro-scopes], in the size ranges of millimeters, microns and nanometers. In particular, the need arises because of the increasing miniaturization and complexity of integrated circuits and cellular studies in the field of bioengineering. However, accurate size measurements cannot be made with a microscope unless it has first been calibrated with a magnification reference “standard” (standard is defined as a device having a verifiable value). The reliability of the information obtained from microscope image measurements depends on the accuracy of the microscope's magnification calibration. Any error in the magnification calibration of a microscope is a component of the total size measurement error and frequently is the predominant component. The magnification scale M of the microscope is determined by the ratio. M=L′/L. L is the size of the test object (measurement reference) used for the magnification calibration; L′ is the size of the same test object realized in the microscope image (the magnified image of the object). The methods of magnification calibration are known B. B. Martinov, “Problems of Measurements of Linear Sizes of Relief Sub Micron Structures on Raster Electronic Microscopes” preprint #501IOFAN.M, 1990, page 18, assuming use of a line-width standard (magnification reference) as the test object. Thus, it is considered that the nominal size L of the magnification reference is known with an adequate accuracy. However, obtaining the exact pitch value L′ from the microscope's image of the test object pattern appears to be a problem in this case. The fact is that a ratio between an object and its image is rather nontrivial. In any kind of microscopy the image is only similar to the object, but is never an exact copy. In particular, there are no universal rules, according to which it would be possible to specify points on the image corresponding to object edges (the distance between which is the image size L′). This is the reason for significant errors in the value of L′ and for microscope magnification calibration errors as a whole. In B. B. Martinov's “Problems of Measurements of Linear Sizes of Relief Sub Micron Structures on Raster Electronic Microscopes” preprint #501IOFAN.M, 1990, page 18, the results of a practical application of this method are given. The calibration error of the Scanning Electron Microscope (SEM) was calculated by the authors as 2.6% in one case and as 5.1% in the other case. This is not an acceptable magnification error for a SEM used to perform accurate measurements. Such SEMs require their magnification error to be less than 0.5%. The use of a pitch magnification reference material as a standard for magnification calibration M. T. Postek, Critical Issues in Scanning Electron Microscopes Metrology, Journ. Of Research of the National Inst. of Standards & Technol., Vol. 99, No. 5, October 1994, pp. 658–660 provides significant advantages in the precision of a microscope's magnification calibration. As the pitch reference contains several or many repeatable identical features (lines or stripes). Independent of the type or model of microscope being calibrated, these patterned lines will appear to be identical to each other. This strongly facilitates evaluation of the pitch value of such structures present in the microscope image: the distance between any equivalent points of adjacent stripe pattern features in the image can be considered as the pitch value. Such points can be established or noted by using the maxima or minima of brightness in the video signal, any repeated characteristic features on the videosignal slopes, etc. In order to implement such a method, firstly it is necessary to create and certify the indicated pitch magnification references with a known accuracy (creating a standard), there by correlating their nominal pitch size to an absolute size scale. Both of these problems are not simple. According to M. T. Postek, Critical Issues in Scanning Electron Microscopes Metrology, Journ. Of Research of the National Inst. of Standards & Technol., Vol. 99, No. 5, October 1994, pp. 658–660, for these reasons up to the present time only two pitch magnification references have been certified to be used as SEM magnification standards. They have the required characteristics and have been created with nominal pitch values less than 1 micron [NIST Standard Reference Materials SRM484 and SRM-2090]. Such magnification references are unique, expensive and not readily available for most users. In particular, pitch magnification reference SRM-2090 contains 8 separate parallel line structures, the pitch value between adjacent line features is about 200 nanometers. It is known that optical diffraction methods can be used to provide highly precise and accurate pitch certifications of a diffraction grating. However, one is not able to apply these techniques to the certification of NIST SRM-484 and SRM-2090 because of the small number of repeated lines. For these purposes the authors U.S. Pat. No. 5,822,875 had to provide a special unique measuring environment that protected the SEM from vibrations and contained a precision stage with a laser inteiferometer, operating under computer control. The achievable accuracy of measurement certification is not given by the authors [U.S. Pat. No. 5,822,875]. Previous inventions attempted to improve the accuracy of SEM measurements, but had a number of drawbacks. A major issues that these inventions failed to provide a universal solution for accurate magnification calibration of image magnifying instruments. More specifically, the invention described in the patent “Scanning electron microscope ruler and method”, U.S. Pat. No. 5,822,875, is restrictive because it requires fabrication of precision features on the integrated circuits to be inspected. This precludes the method from being readily available to a wide number of microscope users. More importantly, the ruler method described in the patent U.S. Pat. No. 5,822,875 did not provide for the highest degree of precision and accuracy. This degree is achievable in our proposed invention. In the patent, “Apparatus and method for measuring length in scanning particle microscope” U.S. Pat. No. 4,677,296, the propose measuring references provided an inferior degree of magnification calibration because they cannot be used as an absolute magnification standard. In addition they do not incorporate the necessary algorithms to determine the statistical measurement errors. Thus, the degree of accuracy of the magnification measurement is unknown. In contrast, our proposed invention enables the user to precisely determine the magnification errors in the microscope. Finally, the invention described in patent, “High precision calibration and feature measurement system for a scanning probe microscope” U.S. Pat. No. 5,825,670 is measuring instrument specific, and has limited use because it requires direct control of the SEM‘s’ scan drive circuitry. Also, it cannot be applied to non-scanning image magnifying instruments. In contrast, our proposed invention can be universally applied to any image magnifying device. The primary object of this invention is to provide a magnification reference with a high degree of pitch accuracy such that it may be employed as a magnification standard. Another object is to provide for the use of the set of algorithms for use with any magnification references-such that the references are composed of equally spaced, parallel line patterns. Another object of the invention is to provide for the use of our proposed magnification reference to further increase the accuracy of the magnification calibration. Another object of the invention is to provide for further increasing of the magnification calibration accuracy on the basis of using a procedure for cutting off the intensity distribution (or videosignal), which is realized in the microscope's image, and subsequent calculation of the set of the “center of mass” positions x0 of the formed “islands”. Another object of the invention is to provide a set of software algorithms for scanning image magnifying instruments (e.g. scanning electron microscopes, conical scanning microscopes, tunneling microscopes and atomic force microscopes). Another object of the invention is to provide a set of software algorithms for non-scanning image magnifying instruments (such as an optical microscope, transmission electron microscopes) where the image is acquired in a digital form, for example using a CCD camera attached to the microscope. Another object of the invention is to provide a precise magnification calibration of a scanning type image magnifying instrument in the horizontal direction of the image plane, such as the horizontal direction of an SEM's scan. A further object of the invention is to provide a precise magnification calibration of a scanning type image magnifying instrument in the vertical direction of the image-plane such as the vertical direction of an SEM's scan. Another object of the invention is to provide an independent calibration in the horizontal and vertical directions of the image plane for scanning type image magnifying instruments. In keeping with these objects and with others which will become apparent hereinafter, one feature of the present invention resides, briefly stated, in a method of precision calibration of a microscope magnification with calculating a magnification scale as a quotient obtained when the image size of a test object viewed or collected with the microscope is divided by the true test object size, comprising the steps of obtaining a magnification reference by taking a diffraction grating with a tested pitch value as the test object; distributing a brightness level between 30–70% amplitude of the video signal image of a diffraction grating obtained in the microscope; calculating the position of the video signal “center of mass” for each of the formed “islands” of the brightness distribution; considering an average distance between neighboring “center of mass” as being a grating pitch in a microscope image of the object; and recognizing that a magnification scale of the microscope is a result of a division of an average pitch dimension by true grating pitch. The novel features which are considered as characteristic for the present invention are set forth in particular in the appended claims. The invention itself, however, both as to its construction and its method of operation, together with additional objects and advantages thereof, will be best understood from the following description of specific embodiments when read in connection with the accompanying drawings. In accordance with the present invention, a method is proposed to perform accurate magnification calibration for non-scanning and scanning microscopes. The prototype for a magnification reference material that is appropriate for performing a precision calibration procedure and being employed as a standard is presented in this patent. As an example, we shall take the method of magnification calibration which uses a special calibration reference having a repeated pitch pattern (pitch standards) 2. This example is based on: 1. The magnification reference containing a multiple of repeatable features. 2. The features in the pattern being of equal-size and spacing. 3. The pattern having parallel feature edges in at least one orthogonal direction. 4. The magnification reference is oriented or positioned for viewing with a microscope in such a manner that the resulting image viewed in the microscope will appear to have feature edges of the magnification reference parallel to the vertical [y] microscope image axis. Performing the procedures specified herein achieves the magnification calibration of the instrument to a high and known accuracy. If a diffraction grating (DG) with pitch T is chosen as the object by which the calibration of a microscope's magnification is to be carried out, the magnification scale M is derived from the ratio: M=T′/T, where T′ is the DG pitch measured on the image (picture) of the object formed by the microscope. The error of the magnification scale ΔM is composed of the error in the nominal value of the DG pitch—ΔT and the error ΔT′ of the pitch measurement on the microscope's image of the DG. High calibration accuracy is achieved when it is possible to decrease these errors to acceptable values. This decrease in the confirmed error of the DG pitch will correspondingly reduce the value of the measurement error in the absolute size scale. This is achieved due to the fact that the test DG consists of a great number of repeating lines that can be attested relative to the average value of its pitch using highly accurate optical diffraction methods. The achievable and achieved accuracy confirmation is evaluated below. Determination of the absolute size scale is accomplished because the wavelength of light used for the grating pitch attestation is usually known in absolute scale and with a relative error of 10−7−10−8 A. N. Zaidel, et al “Tables of Spectral Lines”, M. Fizmatgiz, 1962. The diffraction angle measurement error in the standard optical goniometer is 3–5 arc sec N. P. Gvozdeva, et al “Applied Optics and Optical Measurements, M., Machine Building 1976, which determines the total confirmed error of the average nominal pitch value of the DG ΔT/T is about (3–5)·10−5. Let us estimate the probable value ΔT′. It was experimentally shown that when the standard algorithm using equivalent points of adjacent DG lines (this is obtained from the microscope image) is applied, the standard error of a single measurement is significant. We recommend the measurement algorithm using the “center of mass” approach. The advantages of such an approach will become clear. According to the results of the experiments we carried out, we will consider that the average nominal pitch of the DG we used is 470.72±0.01 nm, and it was obtained from the optical diffraction experiment. The SEM magnification setting was 50,000×. The SEM field of view at such magnification is less than 2 microns. On the SEM screen (or in the SEM picture) 4 DG lines and 3 spaces between them (3 pitch intervals) can be imaged in one line of the image frame. We used a 512 horizontal scan line image frame. The number of pitch measurements in one frame is 1536. If the recommended “mass center” algorithm is used for calculation of the pitch value, the standard error of a single measurement will decrease by up to 6 nanometers. In this case the variation (the standard error) of the mean measured pitch value calculated by averaging over 1536 independent measurements is 0.15 nanometers or 3*10−4 of the nominal pitch value. Mean value of T′ and root-mean-square deviation (RMS) are calculated in accordance with formulae: T ′ = ∑ i = 1 n T i ′ n RMS = 1 n · ∑ i = 1 n ( T i ′ - T ′ ) 2 where n is the number of “islands” in the frame; T′i are the individual pitch values in the microscope's image. From the above, it is seen that the error ΔT′ is the dominant error, and strongly influences the total magnification calibration error. Notice that the same procedure can also be successfully applied to scanning electron microscopes operating at lower magnification in order to accurately determine their magnification scale. At lower magnification, the number of DG lines in each image is proportionally increased and the number of pitch patterns intersected by each horizontal scan line is increased. This results in a decrease of the principle source of error. To calibrate microscopes at higher than 50,000× magnification, it is recommended that a DG with a smaller nominal pitch value be used to provide a higher calculated statistical measurement reliability for the average pitch measurement from the SEM image. A description of achieving our magnification calibration reference follows. The calibration instrument used was a Cambridge Instruments SEM model “Stereoscan-360”. A holographic diffraction grating was used as a magnification reference. The length of the pattern covered by the DG was greater than 1.5 centimeters. The approximate DG pitch value was 0.470 nanometers, and the total number of DG lines was greater than 30,000. The calibration of the DG pitch was performed using a model FC-5 goniometer with scale division intervals of I arc sec [4]. A low-pressure mercury lamp was used as a light source. The diffraction angle measurements were performed using the λ=435.835 nm wavelength mercury line. A set of 30 measurements of the diffraction angle were repeated several times. The results were averaged and the standard mean deviation was calculated. The measured diffraction angle was 67°48′11″±8″ and the average value of the diffraction grating pitch was T (470.72i0.01) nm (1 sigma). The relative error of the pitch value ΔT/T was 2*10−5. The substrate containing the diffraction grating was placed into the SEM chamber and its focused image was collected in digital form. The SEM's operating conditions were: accelerating voltage was 20 kV, the probe current was 120 pA, a coaxial semiconductor backscattered electron detector was used, the frame time was 50 seconds, the number of pixels in the frame was 512×512, the number of gray levels per pixel was 256, and the magnification reading of the digital indicator was 50,000. Processing of the SEM images was carried out using our algorithm on a separate PC to calculate the pitch of the diffraction grating. More than 1500 individual pitch values (in pixels) were used, their standard statistical processing led to the average pitch value T′=120.93±0.07 pixels (1 sigma). The width of the SEM screen Ls was 10 centimeters (512 pixels). the image of the DG was magnified M times: M = L s 512 × T ′ T = 50177 The relative error ΔM/M may be calculated by the ratio:ΔM/M={(ΔT/T)2+(ΔT/T)2}1/2 After substitution of the corresponding values into the above formula we obtain:ΔM/M=6×10−4. Thus, the result of the calibration procedure, i.e. the fixed value of M can be written as:M=(50.18±0.03)×103 These calculations show that the magnification calibration error in the ‘x’ direction on this example results in an error that is less than 0.1% which confirms the fact that we are proposing a high precision method for magnification calibration. Scanning imaging instruments typically require independent magnification calibration in the horizontal and vertical directions of the microscope's image. The example presented above describes a method for calibration of the microscope's image in the horizontal direction of the image plane. To perform a magnification calibration in the vertical direction on the image plane, the magnification reference is rotated 90 degrees such that the edges of the repeating line patterns are parallel with the top and bottom edges of the microscope's image. The image of the magnification reference is then collected and stored in a 512×512 pixel digital format. A series of up to 512 virtual, vertical line segments of the stored digital image are processed as described in the horizontal direction magnification calibration procedure above. These calculations are performed on the storage image data and enable up to 512 independent sets of pitch measurements to be made on each image. The procedure that is then followed is equivalent to the procedure described above for precision magnification calibration and results in independent magnification calibration of the microscope's image in two orthogonal directions (vertical and horizontal). In accordance with the inventive method of achieving a magnification calibration of a microscope and the like, the steps of the method can include selecting a holographic diffraction grating with a nominal pitch size appropriate for a magnification range to be calibrated; selecting a substrate on which a number of grating lines has been sufficiently reproduced; using an optical goniometer with an angular measurement accuracy of 1 arc second; selecting 435.835 nm wavelength mercury line; performing a series of measurements at several locations on a grating; calculating a standard mean deviation from these measurements; calculating a diffraction angle obtained with said goniometer and mercury line; and calculating a diffraction pitch using Bragg's formula: λ=2d*sin θ. It will be understood that each of the elements described above, or two or more together, may also find a useful application in other types of methods and constructions differing from the types described above. |
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claims | 1. A system for evaluating reliability of a component, where said system evaluates credibility of a constituent information and a conformity of a candidate component to environmental regulations, for composing a product, comprising:an input unit;an output unit;a processing unit; anda memory unit,wherein the memory unit comprises:a component basic information table which retains identification information that distinguishes each item of a component;a constituent information table which retains information on constituents contained in the component and a content of each of the constituents; andan environmental qualification information table which retains information on constituents contained in the component for each item and a investigation result of conformity of the component to environmental regulations by establishments, andwherein the processing unit comprises:means which accepts selection of a target component through the input unit;means which accepts identification information that classifies components similar to the target component through the input unit;means which extracts a component similar to the target component based on the identification information of classified components similar to the target component by referring to the component basic information table and the constituent information table, and calculates constituents contained in the similar component and a content of each of the constituents;means which calculates each of the constituents contained in the target component and the content of each constituent, based on the content of each constituent of the component similar to the target component, which is obtained from the constituent information table, the content of each constituent being weighted based on a number of establishments that investigated the similar component, which is obtained from the environmental qualification information table;means which evaluates credibility of the evaluation of the conformity to environmental regulations based on a difference between the content of the constituent in the target component and the content of the constituent in the similar component, the constituent being used for the evaluation of the conformity to environmental regulations among the constituents contained in the target-component; andmeans which evaluates a conformity of the target component to environmental regulations by using the number of establishments that have already investigated the similar components and a total number of the similar components with a result of a qualification “pass ”, which are obtained from the environmental qualification information table;wherein the output unit outputs the result of the conformity/credibility evaluation. 2. A system according to claim 1, wherein the processing unit comprises:means which accepts selection of a status of the target component through the input unit;means which makes a re-investigation request attached with the result of investigation of the target component to a manufacturer of the component when the re-investigation is selected as the status;means which makes a qualification request to a component final confirmation personnel when adoption is selected as the status; andmeans which returns to the input unit of basic information of the target component when non-adoption is selected as the status. 3. A system according to claim 1, wherein the component similar to the target component is extracted referring to the component basic information table by searching for the similar component based on the identification information of the target component according to a definition of the similar component. 4. A method for evaluating reliability on component, which evaluates credibility of an evaluation result of conformity of a component to environmental regulations, comprising the steps of:calculating constituents contained in a target component and a content of each of the constituents referring to a constituent information table based on identification information of the target component and evaluating conformity of the constituent to environmental regulations based on the content;extracting a component similar to the target component referring to an environmental qualification information table and calculating constituents contained in the similar component and a content of each of the constituents; andevaluating credibility of evaluation of the conformity to environmental regulations based on a difference between the content of the constituent in the target component and the content of the constituent in the similar component, the constituent being used for the evaluation of the conformity to environmental regulations among the constituents contained in the target component. |
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047284820 | claims | 1. A method for inservice inspection of a predetermined area of the wall of a pressurized water nuclear reactor pressure vessel wherein said nuclear reactor has a substantially cylindrical pressure vessel wall, with a plurality of inlet and outlet nozzles therein, and a generally cylindrical vertically oriented core barrel having an upper flange disposed within said pressure vessel, with an annular chamber formed between said pressure vessel wall and said barrel, said upper flange resting upon a ledge about the inner periphery of the upper section of said pressure vessel wall, comprising: providing access to said annular chamber through the flange of said core barrel, while said core barrel is disposed within said pressure vessel; inserting a means for inspecting said predetermined area through said access; positioning said inspecting means in proximity to said predetermined area of said pressure vessel wall to be inspected; and inspecting said predetermined area by said inspecting means while said barrel remains disposed within said pressure vessel. providing access to said annular chamber through the flange of said core barrel, while said core barrel is disposed within said pressure vessel, and wherein the upper flange of said core barrel has a plurality of apertures therethrough which communicate with said annular chamber and a removable plug within each said aperture, by removing at least one of said plugs; inserting a means for inspecting a weld, present in the wall of the pressure vessel, through said aperture with said plug removed; positioning said inspecting means in proximity to said weld in said pressure vessel wall to be inspected; and inspecting said weld by said inspecting means while said barrel remains disposed within said pressure vessel. 2. The method for inservice inspection as defined in claim 1 wherein the upper flange of said core barrel has a plurality of apertures therethrough which communicate with said annular chamber and a removable plug within each said aperture, and wherein access to said annular chamber is provided by removing at least one of said plugs. 3. The method for providing inservice inspection as defined in claim 1 wherein access to said annular chamber is provided by forming an aperture through the upper flange of said core barrel, which aperture communicates with said annular chamber. 4. The method for providing inservice inspection as defined in claim 1 wherein said means for inspecting said predetermined area comprises means for ultrasonic testing of said predetermined area. 5. The method for providing inservice inspection as defined in claim 1 wherein said means for inspecting said predetermined area comprises means for visual examination of said predetermined area. 6. The method for inservice inspection as defined in claim 1 wherein said predetermined area includes a weld in said pressure vessel wall. 7. The method for inservice inspection as defined in claim 6 wherein said means for inspecting said weld comprises means for ultrasonic testing of said weld. 8. The method for inservice inspection as defined in claim 1 wherein a nuclear core is present within said core barrel. 9. A method for inservice inspection of predetermined welds present in the wall of a pressurized water nuclear reactor pressure vessel wherein said nuclear reactor has a substantially cylindrical pressure vessel wall, with a plurality of inlet and outlet nozzles therein, and a generally cylindrical, vertically oriented core barrel having an upper flange disposed within said pressure vessel, with an annular chamber formed between said pressure vessel wall and said barrel, said upper flange resting upon a ledge about the inner periphery of the upper section of said pressure vessel wall, comprising: 10. The method for inservice inspection as defined in claim 9 wherein said means for inspecting said weld comprises means for ultrasonic testing of said weld. 11. The method for inservice inspection as defined in claim 10 wherein a nuclear core is present within said core barrel. |
claims | 1. A boiling water type nuclear reactor use control rod which comprises a tie rod having substantially cross shaped cross section, a plurality of sheathes having substantially U shape cross section and attached to the respective sides of the tie rod, a plurality of neutron absorption rods disposed inside the respective sheathes, a handle which is disposed at one ends of the sheathes in the axial direction of the control rod and a lower portion supporting plate or a dropping speed limiter which is disposed at the other ends of the sheathes in the axial direction of the control rod, wherein at least one of the handle, the lower portion supporting plate and the dropping speed limiter is provided with a sliding structural body which is constituted by a pin, a pin hole into which the pin is inserted and a roller which rotates around the pin in the axial direction, and at least one groove is sized to promote water flow in a clearance between the pin and the pin hole and is arranged adjacent the clearance. 2. A boiling water type nuclear reactor use control rod according to claim 1 , wherein the at least one groove is provided either at the outside or at the inside of the clearance. claim 1 3. A boiling water type nuclear reactor use control rod according to claim 2 , wherein the at least one groove is provided at an upstream axial side and at a downstream axial side. claim 2 4. A boiling water type nuclear reactor use control rod according to claim 3 , wherein the at least one groove is formed so as to extend near to end portions of the pin. claim 3 5. A boiling water type nuclear reactor use control rod according to claim 4 , wherein the handle is provided with an opening near the end portions of the pin. claim 4 6. A boiling water type nuclear reactor use control rod which comprises a tie rod having a substantially cross shaped in its cross section, a plurality of sheathes having substantially U shape in its cross section each being attached to the respective four sides of the tie rod, a plurality of neutron absorption rods disposed inside the respective sheathes, a handle disposed at the upper ends of the sheathes and a lower supporting plate or a dropping speed limiter disposed at the lower ends of the sheathes, wherein each of four wings which constitute the handle is provided with a guide use roller constituted by a pin, a pin hole into which the pin is inserted and a roller which rotates in the axial direction around the pin, and a groove is sized to promote water flow in a clearance between the pin and the pin hole and is arranged at two positions adjacent the clearance. 7. A boiling water type nuclear reactor use control rod according to claim 6 , wherein at least one of the lower portion supporting plate and the dropping speed limiter is provided with a second guide use roller constituted by a pin, a pin hole into which the pin is insertably arranged and a roller which rotates in the axial direction around the pin, and a groove which promotes water flow in a clearance between the pin and the pin hole is provided at two positions adjacent to the clearance of the second guide use roller. claim 6 8. A boiling water type nuclear reactor use control rod which comprises a tie rod having a substantially cross shaped in its cross section, a plurality of sheathes having substantially U shape in its cross section each being attached to the respective four sides of the tie rod, a plurality of neutron absorption rods disposed inside the respective sheathes, a handle disposed at the upper ends of the sheathes and a guide use roller which is provided at the handle and permits sliding of the control rod in an axial direction thereof, wherein a groove is sized to promote water flow in a clearance between the pin and the pin hole in the guide use roller and is provided adjacent the clearance at an upstream axial side and at a downstream axial side so as to communicate with the clearance. |
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description | Embodiment of the present invention will be described below in detail, referring to the accompanied figures. Although the reactor cores of electric power of 1350 MW class are described in the following embodiments, the output power capacity is not limited to 1350 MW. It should be recognized that the present invention may be applied to the reactor cores having the other output power capacity by changing number of the fuel assemblies. (First Embodiment) A first embodiment of the present invention will be described, referring to FIG. 1 and FIG. 7 to FIG. 12. FIG. 1 is a cross-sectional plan view showing the first embodiment of a reactor core having an electric output power of 1356 MWe. FIG. 1 shows 504 fuel assemblies 1; and 157 control rod drive mechanisms 2 each of which operates three large-diameter control rods to be inserted into three fuel assemblies, respectively. FIG. 7 shows the cross section of the fuel assembly lattice. In a channel box 3, fuel rods 4 of 10.1 mm diameter are arranged in a regular triangular configuration with a 1.3 mm gap between the rods to form an equilateral hexagonal assembly having 12 fuel rod rows. In the central portion of the fuel assembly, a guide tube 6 to insert the large-diameter control rod 5 thereinto is disposed in the region having an area equivalent to 3 fuel rod layers, that is, an area equivalent to 19 fuel rod unit lattice cells. The large-diameter control rod is formed of an absorption rod of a stainless steel tube filled with B4C. Further, the large-diameter control rod has a follower portion in the top end portion, the follower portion being made of carbon which is a substance having a slowing-down power smaller than that of light water. FIG. 8 shows an arrangement of fuel assemblies under the equilibrium core state. Each of the numerals written in each of the fuel assemblies 1 indicates a period staying in the reactor core by cycle numbers. The 5 cycle fuels staying in the reactor core for the longest period are loaded in the outermost periphery of the reactor core where the neutron importance is low. The fuels of 1 cycle staying period in the reactor core having the highest neutron infinite multiplication factor are loaded in the outer region of the reactor core in the inner side of the outermost periphery to flatten the power distribution in the radial direction of the reactor core. In the inner region of the reactor core, the fuels of 2 to 4 cycle staying periods in the reactor core are distributively loaded to flatten the power distribution in the radial direction of the reactor core. FIG. 9 shows an orifice distribution in the equilibrium reactor core state, and the numeral written in the fuels indicates difference in opening degree of an orifice placed in the fuel supporting portion, and there are two regions for the orifice opening degree. The orifice diameter in the reactor outermost peripheral region (number 1) where the fuel assembly power is small is smaller than the orifice diameter in the inner region. FIG. 10 shows the axial distribution of fissionable plutonium enrichment averaged over the horizontal cross section of the fuel assembly for the equilibrium reactor core. Therein, the uranium to be added with plutonium is the depleted uranium. The height of the reactor core is 70 cm, and the reactor core is divided three regions at the levels of 20 cm and 49 cm from the bottom end of the reactor core, and the fissionable plutonium enrichments are 19 wt %, 0 wt % and 19 wt %, respectively, and the average fissionable plutonium enrichment is 11.1 wt %. Further, depleted uranium blankets having heights of 20 cm and 25 cm are attached to the top and the bottom of the reactor core portion, respectively. FIG. 11 is a horizontal cross-sectional view showing the 19 wt % fissionable plutonium enrichment region of the fuel assembly. The fissionable plutonium enrichments are three kinds of 19.1 wt %, 18.5 wt % and 17.5 wt %, and the average enrichment is 19 wt %. FIG. 12 shows the distributions of core-average output power and core-average void fraction in the axial direction of the reactor core. The core-average void fraction is 60%, and the mass steam quality at reactor core exit is 32 Wt %. Operation of the present embodiment will be described below. By the combination of the regular triangular lattice closed-compact hexagonal fuel assembly having a gap between rods of 1.3 mm, the core-average void fraction of 60% and the large-diameter control rod, an effective water-to-fuel volume ratio of 0.27 was attained, and an in-core breeding ratio of 0.87, a blanket breeding ratio of 0.14 and a total breeding ratio of 1.01 were realized. That is, in the present embodiment, by reducing the effective water-to-fuel volume ratio from nearly 2.0 in the existing reactor to 0.27, the light water reactor having the breeding ratio of 1.01 is realized. The output power of the present reactor core is 1350 MWe which is equal to that of the existing ABWR, and the circumscribed radius of the reactor core is 2.9 m which is nearly equal to that of the ABWR. The height of the reactor core is 70 cm, and the blankets having heights of 20 cm and 25 cm are attached to the top and the bottom of the reactor core to form a short-length fuel assembly. However, since the fuel rods are closely packed, the total length of the fuel rod is nearly equal to that of ABWR fuel and the MCPR is 1.31 which sufficiently satisfies the thermal design standard value of 1.24. Because of the short-length fuel rods having the 70 cm reactor core portion, the plutonium inventory converted to the amount of fissionable plutonium per 1000 MWe output power is as small as 6.0 tons though the fuel rods are closely packed. Even including the period of plutonium staying outside the reactor core such as fuel reprocessing, the plutonium inventory is less than 10 tons per 1000 MWe. From the above reason, in the present embodiment having the breeding ratio of 1.01, using fissionable plutonium of 15 thousands tons and depleted uranium of 15 million tons produced from uranium reserves of 15 million tons in the world, 1500 units of 1000 MWe reactors can be continued to be operated for 10 thousands years and accordingly the system of long-term stable energy supply can be established. In the present embodiment, in regard to the height direction of the fuel assembly, there are the portions having the fissionable plutonium enrichment of 19 wt % in the upper and the lower positions, and the middle portion between them is formed of depleted uranium not containing the fissionable plutonium. When the output power is increased or when the core coolant flow rate is decreased, the steam void fraction in the reactor core is increased. At that time, the power distribution in the upper portion of the reactor core is swung to the middle region of the reactor core where the fissionable plutonium enrichment is not contained. Thereby, the negative void coefficient is inserted. Further, in the present embodiment, since the mass steam quality at the reactor core exit is 32 wt %, and accordingly all the coolant does not become vapor and the coolant can be maintained in a two-phase state even when an abnormal transient event occurs. Therefore, similar to the existing BWR, the radioactive substances accumulated in the reactor core such as corrosion products are enclosed in the reactor core by evaporation operation of boiling and prevented from being transported to the turbine side. From the above reason, the BWR of the present embodiment is capable of cope with the long-term stable energy supply under the degree of safety comparative with that of the fuel burning-only light water reactor under operation now and using the pressure vessel having a size nearly equal to that of the ABWR under construction now. The BWR of the present embodiment outputs the amount of power equal to that of the ABWR, and can attain burn-up of 65 GWd/t. A void coefficient of the existing BWR under operation now (a value at present) is xe2x88x927.0xc3x9710xe2x88x924xcex94k/k/% void. The value for the present embodiment is designed to be xe2x88x920.5xc3x9710xe2x88x924xcex94k/k/% void of which the absolute value is smaller than the value at present. As the result, the thermal margin to an event of increasing pressure or to an event of decreasing coolant temperature is relatively large. From the above reason, the BWR reactor core of the present embodiment has safety margins for various kinds of transient events which are larger than those of the existing BWR under operation now. In the present embodiment, the large-diameter control rod having an outer diameter larger than that of the fuel rod is employed as the absorption rod. By employing the large-diameter control rod, the mechanical strength of the control rod can be increased and accordingly bending and buckling of the control rod can be suppressed when the control rod is inserted or withdrawn. Furthermore, by using the large-diameter control rod, number of absorption rods per fuel assembly can be reduced, and accordingly the control rod can be easily manufactured to reduce the manufacturing cost. According to the present embodiment, by combination of the closed-packed hexagonal fuel assembly, the large-diameter control rod and the core-average void fraction of 60%, the breeding ratio of 1.01 can be realized by the fuel enriched by adding the fissile PU of average 11.1 wt % to the depleted uranium, and 1500 units of 1000 MWe reactors can be operated for 10 thousands years using the uranium reserves of 15 million tons in the world, and accordingly the long-term stable energy supply can be established. Further, since the diameter of the pressure vessel, the operating conditions such as output power and the used materials are the same as those of the existing BWR under operation, the electric power generation cost can be suppressed to the same level as that of the existing BWR even though the performance is largely progressed. Further, by employing the large-diameter control rod, maintaining the negative void coefficient by the axial fuel distribution and suppressing the mass steam quality to nearly 32 wt %, the safety margin can be secured in the same level as that of the existing BWR by maintaining the evaporating function by boiling to enclose the radioactive substances in the pressure vessel. In the present embodiment, aiming at the long-term stable energy supply, the description has be made on the construction, the operation and the effects of the fuel which is enriched by adding plutonium to depleted uranium produced as the residue at manufacturing enriched uranium used for the existing light water reactors. However, the same or more effects can be obtained by the fuel enriched by adding plutonium to natural uranium or the degraded uranium recovered from used fuel or the low enriched uranium instead of the depleted uranium. In this case, the fissile PU enrichment can be reduced by 0.5 wt % or more compared to the case of using the depleted uranium due to increase in an amount of uranium-235 contained in the fuel. As a result, the breeding ratio to the fissile PU can be increased by nearly 3% or more, and the void coefficient can be made negative. In addition, since the Pu inventory can be reduced, number of the RBWRs capable of being operated can be further increased. Although the void coefficient is negative in the present embodiment, the power coefficient including the Doppler coefficient can be made negative even if the void coefficient is 0 or slightly positive. According to the study of the inventors of the present invention, it has been shown from an evaluation result on the safety that the negative or positive void coefficient is essentially no problem if the power coefficient is negative. Therefore, the thermal margin can be increased by increasing the length of the reactor core portion. Further, the breeding ratio can be increased by narrowing the gap between the fuel rods. Although only the fuel enriched by adding Pu to uranium has been described in the present embodiment, the other actinides can be added together with Pu. In this case, since the RBWR is high in the average energy of neutrons, plutonium is hardly converted to actinides having higher mass numbers and at the same time the actinides can be eliminated by nuclear fission reaction. Furthermore, in the present embodiment, there are the portions having the same fissionable plutonium enrichment in the upper and the lower portions, and the middle portion between them is formed of depleted uranium not containing the fissionable plutonium. However, it is not always necessary that the fissionable plutonium enrichments in the upper and the lower portions is equal to each other. Further, the depleted uranium region is arranged in the position slightly higher than the middle position of the reactor core in the present embodiment, but it is not limited to that position. The values of axial power peaking can be made equal by combining the fissile PU enrichments in the upper and the lower portions and the position of the depleted uranium region. (Second Embodiment) A second embodiment of the present invention will be described below, referring to FIG. 13. The present embodiment is a reactor core of an electric power output of 1356 MWe, and has further shortened fuel assemblies. The horizontal cross section and the cross section of the fuel assembly lattice of the present embodiment are the same as FIG. 1 and FIG. 7 of Embodiment 1, respectively. FIG. 13 shows the axial distribution of fissionable plutonium enrichment averaged over the horizontal cross section of the fuel assembly for the equilibrium reactor core. Therein, the uranium to be added with plutonium is the depleted uranium. The height of the reactor core is 45 cm, and the reactor core is divided two regions at the levels of 8/12 from the bottom end of the reactor core, and the fissionable plutonium enrichment in the upper region is 13 wt % and the fissionable plutonium enrichment in the upper region is 12 wt %. Further, depleted uranium blankets having heights of 25 cm and 20 cm are attached to the top and the bottom of the reactor core portion, respectively. Operation of the present embodiment will be described below. The construction of the fuel assembly is the same as that of Embodiment 1. By the combination of the regular triangular lattice closed-compact hexagonal fuel assembly having a gap between rods of 1.3 mm, the core-average void fraction of 60% and the large-diameter control rod, an effective water-to-fuel volume ratio of 0.27 was attained, and a breeding ratio of 1.01 was realized. Comparing with Embodiment 1, in regard to the axial direction of the fuel assembly, the present embodiment does not have the depleted uranium region not containing the fissile PU, and the reactor core portion is as short as 45 cm. Further, in regard to the axial direction of the fuel assembly, the fuel assembly is an upper-and-lower two region fuel having different fissile PU enrichments at the levels of 15 cm from the top end of the reactor core, and the fissionable plutonium enrichment in the upper region is 13 wt % and the fissionable plutonium enrichment in the upper region is 12 wt %. On the other hand, when the steam void fraction in the reactor core is increased, the relative increasing amount of void fraction is large in the lower portion of the reactor core having a lower void fraction than the upper portion of the reactor core already reaching the saturation state. As the result, swing of the neutron flux distribution occurs from the upper portion of the reactor core having a higher neutron importance to the lower portion of the reactor core having a lower neutron importance, and thereby the negative void coefficient is inserted. The RBWR of the present embodiment outputs the amount of power equal to that of the ABWR, and can attain burn-up of 45 GWd/t using the pressure vessel having a size nearly equal to that of the ABWR under construction now. In the present embodiment, the same or more effects can be also obtained by the fuel enriched by adding plutonium to natural uranium or the degraded uranium recovered from used fuel or the low enriched uranium instead of the depleted uranium. Further, the other actinides can be added together with Pu. (Third Embodiment) A third embodiment of the present invention will be described below, referring to FIG. 14 to FIG. 17. The present embodiment is a reactor core in which the electric output power is the same as that of Embodiment 1, and number of fuel assemblies and the structure of the fuel assembly are changed from Embodiment 1. FIG. 14 is a cross-sectional plan view showing the present embodiment of a reactor core having an electric output power of 1356 MWe. FIG. 14 shows 609 fuel assemblies 7; and 193 control rod drive mechanisms 8 each of which operates large-diameter control rods to be inserted into three fuel assemblies. FIG. 15 shows the cross section of the fuel assembly lattice. In a channel box 9, fuel rods 4 of 10.1 mm diameter are arranged in a regular triangular configuration with a 1.3 mm gap between the rods to form an equilateral hexagonal assembly having 11 fuel rod rows. At two positions in the fuel assembly, two guide tubes 11 to insert the large-diameter control rods 10 thereinto are disposed in the regions having an area equivalent to 2 fuel rod rows, that is, an area equivalent to 7 fuel rod unit lattice cells. The large-diameter control rod is formed of an absorption rod of a stainless steel tube filled with B4C. Further, the large-diameter control rod has a follower portion in the top end portion, the follower portion being made of carbon which is a substance having a slowing-down power smaller than that of light water. FIG. 16 shows an arrangement of fuel assemblies under the equilibrium core state. Each of the numerals written in each of the fuel assemblies 7 indicates a period staying in the reactor core by cycle numbers. The 5 cycle fuels staying in the reactor core for the longest period are loaded in the outermost periphery of the reactor core where the neutron importance is low. The fuels of 1 cycle staying period in the reactor core having the highest neutron infinite multiplication factor are loaded in the outer region of the reactor core in the inner side of the outermost periphery to flatten the power distribution in the radial direction of the reactor core. In the inner region of the reactor core, the fuels of 2 to 4 cycle staying periods in the reactor core are distributively loaded to flatten the power distribution in the radial direction of the reactor core. FIG. 17 shows an orifice distribution in the equilibrium reactor core state, and the numeral written in the fuels indicates difference in opening degree of an orifice placed in the fuel supporting portion, and there are two regions for the orifice opening degree. The orifice diameter in the reactor outermost peripheral region (number 1) where the fuel assembly power is small is smaller than the orifice diameter in the inner region. The axial distribution of the fissile PU enrichment averaged with the horizontal cross section of the fuel assembly is the same as that of FIG. 10 of Embodiment 1. The area of the region occupied by the control rod in the present embodiment is decreased from one region of 19 fuel rod unit lattice cells to two regions of 7 fuel rod unit lattice cells, but the control rod value is nearly equal to that of Embodiment 1 because the absorption rods are distributively inserted into the fuel assembly. On the other hand, in the present embodiment, number of fuel rods loaded in the reactor core is increased compared to Embodiment 1, and accordingly the average linear power density is reduced to improve the thermal margin. In the present embodiment, by the combination of the regular triangular lattice closed-compact hexagonal fuel assembly, the large-diameter control rod and the core-average void fraction of 60%, an effective water-to-fuel volume ratio of 0.28 is also attained. As the result, the reactor core characteristics are the same as those of Embodiment 1 and the same effect can be obtained. In the present embodiment, the same or more effects can be also obtained by the fuel enriched by adding plutonium to natural uranium or the degraded uranium recovered from used fuel or the low enriched uranium instead of the depleted uranium. Further, the other actinides can be added together with Pu. (Fourth Embodiment) A fourth embodiment of the present invention will be described below, referring to FIG. 18. The present embodiment is a reactor core of which the reactor core performance is improved on the base of the structure of Embodiment 1. The present embodiment is of 1356 MWe electric output power, and the reactor core cross section is the same as that of FIG. 1 of Embodiment 1. FIG. 18 shows the cross section of the fuel assembly lattice. In a channel box 3, fuel rods 4 of 10.1 mm diameter are arranged in a regular triangular configuration with a 1.3 mm gap between the rods to form an equilateral hexagonal assembly having 12 fuel rod rows. In the central portion of the fuel assembly, a guide tube 6 to insert the large-diameter control rod 5 thereinto is disposed in the region having an area equivalent to 3 fuel rod rows, that is, an area equivalent to 19 fuel rod unit lattice cells. Outside of the guide tube, water excluding rods 12 for excluding the moderator between the guide tube and the fuel rods adjacent to the guide tube are arranged. The large-diameter control rod is formed of an absorption rod of a stainless steel tube filled with B4C. Further, the large-diameter control rod has a follower portion in the top end portion, the follower portion being made of carbon which is a substance having a slowing-down power smaller than that of light water. All of the configuration of fuel assemblies in the reactor core, the orifice state and the axial distribution of fissionable plutonium enrichment averaged over the horizontal cross section of the fuel assembly for the equilibrium reactor core are the same as FIG. 8, FIG. 9 and FIG. 10 of Embodiment 1, respectively. In the present embodiment, comparing with Embodiment 1 the effective water-to-fuel volume ratio can be improved and the power peaking in the fuel assembly can be suppressed by excluding the moderator around the guide tube. The reactor core characteristics are the same as those of Embodiment 1 and the same effect can be obtained. In the present embodiment, the same or more effects can be also obtained by the fuel enriched by adding plutonium to natural uranium or the degraded uranium recovered from used fuel or the low enriched uranium instead of the depleted uranium. Further, the other actinides can be added together with Pu. (Fifth Embodiment) A fifth embodiment of the present invention will be described below, referring to FIG. 19. The present embodiment is a reactor core of which the reactor core performance is improved on the base of the structure of Embodiment 1. The present embodiment is of 1356 MWe electric output power, and the reactor core cross section is the same as that of FIG. 1 of Embodiment 1. FIG. 19 shows the cross section of the fuel assembly lattice. In a channel box 3, fuel rods 4 of 10.1 mm diameter are arranged in a regular triangular configuration with a 1.3 mm gap between the rods to form an equilateral hexagonal assembly having 12 fuel rod rows. At three positions in the fuel assembly, three guide tubes 11 to insert the large-diameter control rods 10 thereinto are disposed in the regions having an area equivalent to 2 fuel rod rows, that is, an area equivalent to 7 fuel rod unit lattice cells. The large-diameter control rod is formed of an absorption rod of a stainless steel tube filled with B4C. Further, the large-diameter control rod has a follower portion in the top end portion, the follower portion being made of carbon which is a substance having a slowing-down power smaller than that of light water. All of the configuration of fuel assemblies in the reactor core, the orifice state and the axial distribution of fissionable plutonium enrichment averaged over the horizontal cross section of the fuel assembly for the equilibrium reactor core are the same as FIG. 8, FIG. 9 and FIG. 10 of Embodiment 1, respectively. The area of the region occupied by the control rod in the present embodiment is decreased from one region of 19 fuel rod unit lattice cells of Embodiment 1 to three regions of 7 fuel rod unit lattice cells. Thereby, the absorption rods can be distributively inserted into the fuel assembly, and consequently the control rod value is improved compared to Embodiment 1. The other reactor core characteristics are the same as those of Embodiment 1 and the same effect can be obtained. In the present embodiment, the same or more effects can be also obtained by the fuel enriched by adding plutonium to natural uranium or the degraded uranium recovered from used fuel or the low enriched uranium instead of the depleted uranium. Further, the other actinides can be added together with Pu. (Sixth Embodiment) The present embodiment is a case where the present invention is applied to a squire fuel assembly. FIG. 20 shows the construction of the present embodiment of the fuel assembly. In a channel box 13, fuel rods 14 of 10.8 mm diameter are closely arranged in a regular triangular configuration with a 1.3 mm minimum gap between the rods. In the central portion of the fuel assembly, a guide tube 16 to insert the large-diameter control rod 15 thereinto is disposed in the region having an area equivalent to 2 fuel rod rows, that is, an area equivalent to 7 fuel rod unit lattice cells. The large-diameter control rod is formed of an absorption rod of a stainless steel tube filled with B4C, and the large-diameter control rod has a follower portion in the top end portion, the follower portion being made of carbon which is a substance having a slowing-down power smaller than that of light water. The large-diameter control rods to be inserted into four of the fuel assemblies are operated by one control rod driving mechanism. In the present embodiment, in order to flatten the fuel rod power peaking in the fuel assembly, the fissile PU enrichment of fuel rods facing the channel box and fuel rods facing the guide tube is made lower than that of the other fuel rods. In the present embodiment, by the combination of the regular triangular lattice closed-compact square fuel assembly having the minimum gap between rods of 1.3 mm, the large-diameter control rod and the core-average void fraction of 60%, an effective water-to-fuel volume ratio of 0.34 was attained, and a breeding ratio of 1.01 was realized. In the present embodiment, the same or more effects can be also obtained by the fuel enriched by adding plutonium to natural uranium or the degraded uranium recovered from used fuel or the low enriched uranium instead of the depleted uranium. Further, the other actinides can be added together with Pu. (Seventh Embodiment) A seventh embodiment of the present invention will be described below, referring to FIG. 21 to FIG. 25. The present embodiment is a reactor core in which the electric output power is the same as that of Embodiment 1, and the number of fuel assemblies, the structure of the fuel assembly and the control rod drive mechanism are changed from Embodiment 1. FIG. 21 is a cross-sectional plan view showing the present embodiment of a reactor core having an electric output power of 1356 MWe. FIG. 21 shows 313 fuel assemblies 18; and 313 control rod drive mechanisms 18 each of which operates a large-diameter control rod to be inserted into one fuel assembly. FIG. 22 shows the cross section of the fuel assembly lattice. In a channel box 19, fuel rods 4 of 10.1 mm diameter are arranged in a regular triangular configuration with a 1.3 mm gap between the rods to form an equilateral hexagonal assembly having 15 fuel rod rows. In the central portion of the fuel assembly, a guide tube 21 to insert the large-diameter control rod 20 thereinto is disposed in the region having an area equivalent to 4 fuel rod rows, that is, an area equivalent to 37 fuel rod unit lattice cells. The large-diameter control rod is formed of an absorption rod of a stainless steel tube filled with B4C. Further, the large-diameter control rod has a follower portion in the top end portion, the follower portion being made of carbon which is a substance having a slowing-down power smaller than that of light water. FIG. 23 shows an arrangement of fuel assemblies under the equilibrium core state. Each of the numerals written in each of the fuel assemblies 17 indicates a period staying in the reactor core by cycle numbers. The 5 cycle fuels staying in the reactor core for the longest period are loaded in the outermost periphery of the reactor core where the neutron importance is low. The fuels of 1 cycle staying period in the reactor core having the highest neutron infinite multiplication factor are loaded in the outer region of the reactor core in the inner side of the outermost periphery to flatten the power distribution in the radial direction of the reactor core. The fuels of 2 to 4 cycle staying periods in the reactor core are distributively loaded in the inner region of the reactor core, and one 5 cycle fuel staying in the reactor core is loaded at the center of the reactor core. By doing so, the power distribution in the inner region is flattened. FIG. 24 shows an orifice distribution in the equilibrium reactor core state. The numeral written in the fuels indicates difference in opening degree of an orifice placed in the fuel supporting portion. There are 6 regions for orifice opening degree in total, that is, 5 regions for individual cycles staying in the reactor core shown in FIG. 23 and 1 region for the center of the reactor core. The orifice diameter in the reactor outermost peripheral region (number 5) where the fuel assembly output power is small is smaller than the orifice diameters in the inner region. FIG. 25 shows the axial distribution of the fissile PU enrichment averaged with the horizontal cross section of the fuel assembly. The uranium to be added with Pu is depleted uranium. In the present embodiment, number of fuel assemblies to be loaded in the reactor core is reduced from 504 assemblies of Embodiment 1 to 313 assemblies by increasing number of fuel rods per one fuel assembly to make the fuel assembly large in size, and thereby the reactor core is made small in size. By making the fuel assembly large in size and at the same time by increasing the region occupied by the control rod from the area equivalent to 19 fuel rod unit lattice cells of Embodiment 1 to the area equivalent to 37 fuel rod unit lattice cells, the control rod value is made nearly equivalent to that of Embodiment 1. Further, One unit of the control rod drive mechanism is used for each of the control rods to be inserted into the fuel assembly. In the present embodiment, by the combination of the regular triangular lattice closed-compact hexagonal fuel assembly, the large-diameter control rod and the core-average void fraction of 60%, an effective water-to-fuel volume ratio of 0.27 is also attained. As the result, the reactor core characteristics are the same as those of Embodiment 1 and the same effect can be obtained. In the present embodiment, the reactor core is constructed so that the large-diameter control rod is also inserted into the fuel assembly loaded in the outermost periphery of the reactor core. However, a reactor core may be designed so that the control rod is not inserted into the fuel assembly in the outermost periphery which has a small effect on securing reactor shut-down margin. In the present embodiment, the same or more effects can be also obtained by the fuel enriched by adding plutonium to natural uranium or the degraded uranium recovered from used fuel or the low enriched uranium instead of the depleted uranium. Further, the other actinides can be added together with Pu. (Eighth Embodiment) An eighth embodiment of the present invention will be described below, referring to FIG. 26. The present embodiment is a reactor core in which the electric output power is the same as that of Embodiment 1, and number of fuel assemblies, the structure of the fuel assembly and the control rod drive mechanism are changed from Embodiment 1. The present embodiment has an electric output power of 1356 MWe, and the reactor core is the same as FIG. 21 of Embodiment 7. FIG. 26 shows the cross section of the fuel assembly lattice. In a channel box 19, fuel rods 4 of 10.1 mm diameter are arranged in a regular triangular configuration with a 1.3 mm gap between the rods to form an equilateral hexagonal assembly having 15 fuel rod rows. At two positions in the fuel assembly, two guide tubes 6 to insert the large-diameter control rods 5 thereinto are disposed in the regions having an area equivalent to 3 fuel rod rows, that is, an area equivalent to 19 fuel rod unit lattice cells. The large-diameter control rod is formed of an absorption rod of a stainless steel tube filled with B4C. Further, the large-diameter control rod has a follower portion in the top end portion, the follower portion being made of carbon which is a substance having a slowing-down power smaller than that of light water. All of the configuration of fuel assemblies in the reactor core, the orifice state and the axial distribution of fissionable plutonium enrichment averaged over the horizontal cross section of the fuel assembly for the equilibrium reactor core are the same as FIG. 23, FIG. 24 and FIG. 25 of Embodiment 7, respectively. The area of the region occupied by the control rod in the present embodiment is decreased from one region of 37 fuel rod unit lattice cells of Embodiment 7 to two regions of 19 fuel rod unit lattice cells. Thereby, the absorption rods can be distributively inserted into the fuel assembly, and consequently the control rod value is improved compared to Embodiment 7. The other reactor core characteristics are the same as those of Embodiment 7 and the same effect can be obtained. In the present embodiment, the same or more effects can be also obtained by the fuel enriched by adding plutonium to natural uranium or the degraded uranium recovered from used fuel or the low enriched uranium instead of the depleted uranium. Further, the other actinides can be added together with Pu. (Ninth Embodiment) The present embodiment is a case where the present invention is applied to a squire fuel assembly. FIG. 27 shows the construction of the present embodiment of the fuel assembly. In a channel box 22, fuel rods 23 of 9.8 mm diameter are closely arranged in a regular triangular configuration with a 1.3 mm minimum gap between the rods. In the central portion of the fuel assembly, a guide tube 25 to insert the large-diameter control rod 24 thereinto is disposed in the region having an area equivalent to 4 fuel rod rows, that is, an area equivalent to 37 fuel rod unit lattice cells. The large-diameter control rod is formed of an absorption rod of a stainless steel tube filled with B4C, and the large-diameter control rod has a follower portion in the top end portion, the follower portion being made of carbon which is a substance having a slowing-down power smaller than that of light water. The large-diameter control rods to be inserted into four of the fuel assemblies are operated by one control rod driving mechanism. In the present embodiment, in order to flatten the fuel rod power peaking in the fuel assembly, the fissile PU enrichment of fuel rods facing the channel box and fuel rods facing the guide tube is made lower than that of the other fuel rods. In the present embodiment, by the combination of the regular triangular lattice closed-compact square fuel assembly having the minimum gap between rods of 1.3 mm, the large-diameter control rod and the core-average void fraction of 60%, an effective water-to-fuel volume ratio of 0.34 was attained, and a breeding ratio of 1.01 was realized. In the present embodiment, the same or more effects can be also obtained by the fuel enriched by adding plutonium to natural uranium or the degraded uranium recovered from used fuel or the low enriched uranium instead of the depleted uranium. Further, the other actinides can be added together with Pu. (Tenth Embodiment) A tenth embodiment of the present invention is explained, referring to FIG. 28. The present embodiment is a modification of the eighth embodiment, and it is a core in which the number of fuel assemblies, the construction of each fuel assembly and each control rod drive mechanism are changed, for the same electric output as in the first embodiment. In the present embodiment, the electric output is 1356 MWe, a core cross-sectional view thereof is the same as FIG. 21 of the seventh embodiment. FIG. 28 shows a cross-section of a fuel assembly lattice. Inside a channel box 19, fuel rods 4 of diameter 10.1 mm are arranged in a regular triangular configulation with a gap of 1.3 mm between the fuel rods to form an equilateral hexagonal fuel assembly having 15 fuel rod rows. Inside the fuel assembly, guide tubes 6, in each of which a large-diameter control rod 5 is inserted, are arranged at six locations, and each guide tube 6 is disposed in a region having an area equivelent to two fuel rod rows, that is, an area equivalent to 7 fuel rod unit lattice cells. The large-diameter control rod is formed of an absorption rod of a stainless steel tube filled with B4C. Further, the large-diameter control rod has a follower portion in the top end portion, the follower portion being made of carbon which is a substance having a slowing-down power smaller than that of light water. All of the configuration of fuel assemblies in the reactor core, the orifice state and the axial distribution of fissionable plutonium enrichment averaged over the horizontal cross section of the fuel assembly for the equilibrium reactor core are the same as FIG. 23, FIG. 24 and FIG. 25 of Embodiment 7, respectively. The area of the region occupied by the control rod in the present embodiment is decreased from one region of 37 fuel rod unit lattice cells of Embodiment 7 to six regions of 7 fuel rod unit lattice cells. Thereby, the absorption rods can be distributively inserted into the fuel assembly, and consequently the control rod value is improved compared to Embodiment 7. The other reactor core characteristics are the same as those of Embodiment 7 and the same effect can be obtained. In the present embodiment, the same or more effects can be also obtained by the fuel enriched by adding plutonium to natural uranium or the degraded uranium recovered from used fuel or the low enriched uranium instead of the depleted uranium. Further, the other actinides can be added together with Pu. According to the present invention, by attaining the breeding ratio of near 1.0 or more than 1.0 using the fuel which is enriched by adding plutonium or plutonium to depleted uranium, natural uranium, degraded uranium or low enriched uranium, the depleted uranium, the natural uranium, the degraded uranium or the low enriched uranium can be burned using the plutonium like a catalyst, which can contribute to the long-term stable energy supply. |
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description | The present invention relates generally to a detector used in an electron column system and, more particularly, to a detector which easily detects electrons and secondary electrons resulting from the collision of an electron beam generated by an electron column with a sample, and a method of easily performing detection using the electron column. Conventionally, electrons generated by an electron microscope or an electron beam generator hit a sample where they either induce the reflection of the electron or the emission of secondary electrons. All electrons can be detected by a conventional micro-channel plate detector (MCP), a secondary electron (SE) detector or a semi-conductor detector. When electrons enter these detectors (system) the number of electrons is multiplied through their own structure. For such amplification, a some voltage is applied to a detector or a potential difference generated due to their structure and material is produced. The electric current generated by the electrons undergoing the procedure in above-described detector is amplified by an external amplification circuit. Referring to FIG. 1, an electron beam B emitted from an electron beam generator 100 is projected onto a sample. A conventional detector 10 detects electrons 9 either from the collision of the electron beam B with the sample or secondary electrons emitted from the sample. The electrons 9 are emitted or bounced from the sample—as indicated by arrows and dots—and are then detected by the detector 10. However a conventional detector 10 uses a method for performing amplification to collect data using electrons resulting from the collision of the electron beam with the sample or secondary electrons. The number of electrons resulting from the collision of the electron beam B with the sample and the number of secondary electrons, generated by a conventional electron beam generator, is small. The electric current from these detected electrons has to be amplified before usage. In an electron beam generator such as an electron microscope with high energy or a conventional electron beam generator with low energy typical value for the electric current range from Pico ampere to more than hundreds of Pico amperes depending on the application conditions. Therefore, such a detector (detecting system) is configured to immediately multiply the number of collected electrons. For example, a Micro Channel Plate (MCP) or a Back Scattering Electron Detector (BSED) using a P-N junction has been used. The recent development of an electron beam emission source, such as a micro-column, generates an electron beam with a much higher current at low energy. The number of electrons reaching and interacting with the sample is in the range from hundreds of Pico amperes to Nanoamperes. Compared to a conventional electron column these currents can easily be detected. The present invention has been made keeping the above occurring problems in mind. The objective of the present invention is to provide a cost efficient detector, which is simple in structure and the convenient in usage. Another objective of the present invention is to provide a method of simply detecting electrons by using an electron column. In order to accomplish the above objectives, a detector for an electron column according to the present invention is made of conductive material in a mesh form. This structure is made from an arrangement of one or more wired arrangements or a conductive plate, so that the electron column is placed on a sample and then used. In order to accomplish the above objectives, the method of detecting electrons according to the present invention can perform detection using the above detector or checking sample current using the conductive part of a sample. The method of detecting electrons generated by an electron beam in a column according to the present invention is characterized in that a detector directly detects the electrons and provides the electric current as data about the detected electrons to the outside. In contrast to a conventional electron beam generator, the number of electrons emitted from the source onto a sample by a micro-column is relatively high and hence the number of electrons reflected back from the sample or produced by the emission of secondary electrons is relatively high. Electric currents between about 0.5 nA and several Nanoamperes are measured. Therefore, a reliable detector with a simple structure is used instead of the conventional detector with more a complicated structure. The structure of the detector according to the present invention can be patterned as conductive wiring. The conductive wiring is characterized in that it performs the role of a conventional detector even if the conductive wiring on a path, which does not obstruct the path of the electron beam from the micro-column, is used and it is connected to an external amplifier or a control device. Furthermore, the detection method according to the present invention detects using the detector according to the present invention or utilizes a sample current method. In the detection method according to the present invention the detector is located around a sample and electrons from the sample are detected by the detector, or performing wiring on a sample as the sample current method and directly detecting electrons from the sample using a wire without a separate detector. The detector according to the present invention has a simple structure, since in the detector the structure or processing for multiplying the number of electrons detected is not required, and the manufacturing cost are inexpensive compared to a conventional detector. Furthermore, the detecting method according to the present invention easily detects electrons using the detector according to the present invention or a sample current method. A detector according to the present invention is described below in detail with reference to the drawings. FIG. 2 is a schematic sectional view schematically illustrating the use of a detector, which is an example of the present invention. Electrons emitted from an electron emission source 1 are converted into an electron beam B with a shape given by source lens 3, then deviated by a deflector and finally focused on a sample by a focus lens 6. A detector 20 detects electrons resulting from the collision of the electron beam with the sample as well as secondary electrons from the sample. While the electrons are moving (indicated by arrows) from the sample, the electric current—produced by the electrons 9 hitting the detector (indicated by dots)—is analyzed. This results in data about the sample. FIG. 3 is a schematic perspective view illustrating the detection using detector according to the present invention, which illustrates the arrival of the electron beam B to the sample through the inner space of the detector 20 in the sectional view of FIG. 2. In FIG. 3, the detector 20 is constructed by crossing four conductive wires and the electron beam B is radiated through a space between the conductive wires. FIG. 4 is a schematic perspective view illustrating the positioning of the detector of the present invention above a sample. The figure shows the detector as a mesh formed by wires. The size of each mesh space is larger than the width of the electron beam B. In this case the four wires cross each other and the electron beam passes through a mesh space and is emitted onto the sample. The detector 20 is connected to an external controller via a detector wire 21. The detector according to the present invention is connected by a conductive wire to a controller such as a detector amplifier. It collects electrons resulting from the reflection of an electron beam from a sample or of secondary electrons from the sample, and transmits the signal to an amplification circuit or a controller. The simplest detector structure is a single conductive wire located close to the area to be scanned and moved when the scan area is moved. With increasing number of conductive wires, the detection efficiency increases. For example, when a grid of conductive wires is located over a scan area the detection of electrons is more effective than with just a single wire. In this case the number of electrons from the sample hitting the grid of conductive wires is much increased in comparison to the single wire. It only has to be ensuring that the grid of conductive wires does not obstruct the path of the electron beam passing through. That is, the grid-patterned conductive wires can be used if the number of electrons detected from an electron beam is large regardless of the sample, and thus the detection of electrons from the sample is not obstructed. Nonetheless, if the scan range of the electron beam is broader than the sensitive area of the detector, the possibility that the detector interferes with the electron beam increases since the detector may overlap the scan range of the electron beam. This can be checked by conducting an image test using a test sample. In FIG. 3, the image of the test sample has been previously acquired by the detector of the present invention. The original image is then compared with the image of the detector test. Furthermore, the analysis of the measured electric current of the test sample can perform inspection similar to the image test. A method for analyzing the electric current acquired by the detector and comparing it with the current data of the test sample may be used. If the detector obstructs the image of the sample, the sample or the detector is moved and the image or the current data can be checked. The detector according to the present invention may be formed of conductive wires as described above or may have a structure like a grid mesh shape. It is also possible to form the conductive wires into a planar structure with a circular or a polygonal shape. If required it is possible to form the conductive wires into a three-dimensional shape (for example a pyramid or a cone-shaped structure). It is also possible to use a conductive plate instead of conductive wiring and arrange it in a way so that the conductive plate does not interfere with the projection of the electron beam onto a desired area on the sample, thereby using it as a detector. The method of detecting electrons in an electronic column according to the present invention is conducted by performing the wiring to the outside, amplifying the electric current, and analyzing the information about the image or other samples by checking the amplified values and the variation thereof. In comparison to a conventional detector with a specially shaped structure to which a voltage is applied to the detector and the detector multiplies electrons (increases the number of electrons) and acquires information from detected electrons, the detector of the present invention requires only wiring, and analyzes electrons (current value and variation thereof) detected through the wiring, thereby analyzing data. Another method of detecting electrons in an electronic column system according to the present invention is to perform detection using the wiring of a sample through a conventional sample current method without using a separate detector. The sample current method is to arrange wiring on a sample, amplify the number of electrons detected through the wiring with an extra amplifier, and analyze information related to the detected electrons. This concept pertains to a method of arranging wiring on part of the sample and then performing detection, rather than arranging wiring on the entire sample. FIG. 5 is a diagram illustrating the concept and arrangement of a method of performing detection using the sample current method. As illustrated in FIG. 5, in the case where the arrangement of wiring is known for example, the internal wiring of a semiconductor, the adjacent data lines or gate lines of a Thin Film Transistor (TFT) in a TFT-LCD, or a display with wires arranged cross to each other—the present invention utilizes any piece of wiring as conducting wires of a detector. In FIG. 5 a wire 50 passing closely to an examined sample is used as a detector (part). The wire 50 is connected to an ampere meter A that can analyze the current—such as the intensity of detected current—thereby analyzing and utilizing this data. When an electron beam scans a predetermined scan area 51 a related image is generated as represented in FIG. 6. Furthermore, the current intensity is represented as in FIG. 7 when the electron beam performs line-scanning as in a scan line 52. When the sample is a grounded conductor, the electrons are detected well without the discharge of electrons due to the electron beam If a negative voltage is applied to the sample from the outside, the detection is performed better. If the sample is not a conductor, detection is still possible but charging effects occur. Thus to minimize charging effects it is preferable to perform scans without exposing same part of the sample several times. As the results of the detection in FIG. 5, a detected image is represented using the sample current method as in FIG. 6, and current intensity data is acquired as in FIG. 7. The image of FIG. 6 has been edited and then illustrated for convenience of viewing. The current data of FIG. 7 is actual data. The examination areas chiefly include downward partial peak parts 64, and the current values at the detector represent lower peak parts 62. In FIG. 6, block bar 61 designates the conductive part of the detector shown in FIG. 5, and block bars 63 designates a bar which is a target to be detected. In FIG. 6, although block bars 61 and 63 may be considered similar in brightness, the area of the detector is more clearly viewed in an actual image and the image of the area of the sample to be detected is less clearly viewed than the area of the detector. This is known from the current data of FIG. 7. Block bar 61 corresponds to the detector, represents the current data of lower peak parts 62 and the block bar 63 corresponding to the sample represents upper peak parts 64. Since the number of electrons radiated onto the sample area is lower than the number of electrons radiated directly onto the detector itself, the images is lower bright and the current data value is lower. In FIG. 7, because electrons have negative charges, the current value of the lower portion is higher than that of the upper portion, which means that more electrons are detected in the lower portion. Therefore electrons from the electron beam emitted onto and reflected by a conductor around a detection area or secondary electrons are detected by the conducting parts of the sample. An image or electric current data can be acquired from the sample using the sample current method without using a separate detector. When a negative voltage is applied to the sample parts in the example of FIG. 5, more emitted electrons are detected by the detector, providing a clearer image or a higher electric current further improving the analysis. If a negative voltage is applied to the sample parts the electrons from the column are reflected from the sample parts towards the detector part and are detected more easily by the detector part. If the sample part in FIG. 5 is not a conductor and is cut when the electron beam performs scanning, the electrons emitted from the electron beam are charged at ends spaced apart from each other, so that the charged electrons at ends spaced apart are analyzed as current data, and the sample part is thereby examined. If the sample in the case of FIG. 5 is a conductor, the examination (sample) area and the detection area are exchanged and then detection can be performed. The results of the detection are identical to the case mentioned above. Although a single-type electron column is described, a multi-type electron column can utilize the detector and the detection method according to the present invention. In the multi-type electron column many electron column units—each is a single-type electron column—may be arranged in an n×m matrix and then be used, or may be used in a wafer form. Although single electron columns can have each a detector in entire columns with 1:1 correspondence, the single columns are sequentially used and detection is performed if one detector according to the present invention is used for several columns. When the single columns are operated at specific time intervals, data can be detected only sequentially. If one detector is used for several columns, the detector may be broadly arranged above the sample. If the detector uses a wire type setup, the wire is broadly arranged above the sample. In particular, when detection is performed using conductive wires within a sample, detection can be performed by the sequential operation of respective unit single columns operated rather than specific detection. The detector and detection method according to the present invention can be used in all fields using an electron column. The detector and detection method according to the present invention can be used in all applications that use an electron column, such as an electron microscope, an examination device using an electron beam, and a lithography device using an electron beam. |
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054127017 | abstract | Coatings for zirconium alloy components of nuclear reactor fuel assemblies are described. The coating consists of a metal silicate binder, particles of burnable-poison particles, such as boron carbide, optional graphite particles and an optional rheology-enhancing component. The coating is deposited from a liquid suspension which also includes a polar solvent. |
043449115 | summary | This invention relates to inertial fusion systems and particularly to means for protecting inner structural components of such inertial fusion systems from the x-rays, neutrons, plasma, shock effects, etc. produced by implosion of fusion targets therein. For decades, efforts have been carried out to utilize fusion energy as a source of useful power. Fusion energy should: (1) be an abundant source, (2) be safe, (3) be compatible with the environment, and (4) be technically and economically feasible. Inasmuch as estimated reserves of fusion fuel sources appear to provide the capability of supplying projected electrical energy needs for several hundred years, thus satisfying item 1, efforts have been directed to satisfying items 2-4. Early and currently ongoing efforts are directed to the production of fusion power by the magnetic confinement approach, with more recent efforts also being directed to the inertial confinement approach. With the advent of lasers, early inertial confinement efforts have been directed to the development of laser initiated fusion power plants, as exemplified by U.S. Pat. Nos. 3,624,239 issued Nov. 30, 1971 to A. P. Fraas; 3,723,246 issued Mar. 27, 1973 to M. J. Lubin; and 3,762,992 issued Oct. 2, 1973 to J. C. Hedstrom, wherein a fuel containing target is injected into an implosion or combustion chamber and imploded by one or more laser beams directed into the chamber. More recently, development efforts have also been directed to utilizing ion and electron beams for imploding fusion targets within a chamber, as exemplified by U.S. Pat. Nos. 3,892,950 issued July 1, 1975 to J. R. Freeman et al; and 3,899,681 issued Aug. 12, 1975 to E. H. Beckner et al. With the experimental verification in 1974 and 1975 of the production of neutrons, x-rays, etc. by the implosion of tiny fusion fuel targets via inertial confinement, and with the verification in 1976 that the neutrons thus produced were indeed thermonuclear, efforts have been substantially increased in the field of inertial fusion reactor development. As the result of the experimental efforts in producing fusion neutrons from the tiny targets, it was found that implosion of such resulted in very little damage to the wall surfaces, etc. of the implosion or combustion chamber. However, it is recognized that implosion of larger targets at a selected repetition rate will be necessary to produce useful power by inertial confinement systems, and thus means must be developed to protect the implosion chamber from the x-rays, high-energy neutrons, etc., produced by the implosion of these larger or higher yield targets. At the presently contemplated values of .rho.R (=product of final density and final radius of the imploded fusion target) of 1-2 for laser fusion experiments, approximately 75% of the fusion energy thus produced is in the form of high energy neutrons, with the remainder being primarily x-rays and target debris. A neutron moderating material is thus required to convert neutron kinetic energy to thermal energy. Lithium has been considered as such a moderating material, as exemplified by above-referenced U.S. Pat. No. 3,624,239. A DT fusion reactor must also breed its own tritium, and the only tritium-producing reactions with sufficiently high cross sections to be useful are those involving lithium: EQU .sup.6 Li+n.fwdarw..sup.4 He+T EQU .sup.7 Li+n.fwdarw..sup.4 He+T+n To obtain a T breeding ratio >1.0, the .sup.7 Li reaction is required to offset unavoidable neutron losses; this reaction produces a tritium atom without depleting the neutron population, although said reaction has an energy threshold of 4 MeV and a much lower reaction cross section than the .sup.6 Li reaction. At the same time, it is also necessary to protect the first exposed wall in the interior of the fusion chamber from the debris, x-rays, and high energy neutrons (approximately 25% of the total energy) produced by each microexplosion. To accomplish the protection of the chamber, the breeding of its own tritium, and the dissipation of the heat produced by the conversion of neutron kinetic energy, the so-called "wetted wall" approach has been proposed, as exemplified by above-referenced U.S. Pat. No. 3,762,992, utilizing liquid lithium. The so-called "dry wall" approach has been proposed wherein a sacrificial metal or ceramic liner is placed between the fusion chamber and the blanket which interacts with the x-rays and debris. In addition, the so-called magnetically protected wall uses a solenoid to divert the pellet debris away from the wall into collectors. While these prior approaches have been calculated to provide adequate protection from x-rays and debris, the structures are still subject to damage from high energy neutrons, and only for a period of time, possibly as long as 1-3 years. A need exists by which structural wall protection can be accomplished more effectively and more economically. SUMMARY OF THE INVENTION The invention is directed to means for protecting the first or inner wall of an inertial fusion implosion chamber from high energy neutrons, x-rays, charged particles, and debris produced by the implosion of fusion fuel targets in the chamber, as well as providing a neutron moderating material to convert neutron kinetic energy to thermal energy, and for breeding tritium. This is accomplished in accordance with the invention by providing a blanket within the chamber which utilizes a fluidized wall similar to a waterfall composed of liquid or lithium or of solid pellets of lithium-ceramic. Calculations indicate that the lithium waterfall approach will provide adequate protection of the chamber for about 30 years. Therefore, it is an object of this invention to provide means for protecting the first wall of a fusion chamber from x-rays, neutrons, etc. created by implosion of a fusion fuel target therein. A further object of the invention is to provide in a fusion reaction chamber a blanket which utilizes a fluidized wall for protecting the inner surface of the chamber. Another object of the invention is to provide means for the protection of the inner components of a fusion reaction chamber from x-rays, neutrons, etc. while providing for tritium breeding and for conversion of neutron kinetic energy to thermal energy. Another object of the invention is to provide a lithium waterfall blanket for use in laser fusion implosion chambers. Other objects of the invention will become readily apparent from the following description and accompanying drawings. |
claims | 1. A protection system for a nuclear boiling water reactor, the protection system comprising:a device configured to monitor reactor power;a device configured to monitor reactor pressure;a device configured to determine a power-dependent high reactor pressure setpoint, based on the monitored reactor power; anda device configured to initiate a protection system action when the monitored reactor pressure is greater than the power-dependent high reactor pressure setpoint;wherein the power-dependent high reactor pressure setpoint that corresponds to a value of percent power in an operating domain of the reactor is less than the power-dependent high reactor pressure setpoint that corresponds to 100% reactor power. 2. The system of claim 1, wherein the protection system action is a reactor scram. 3. The system of claim 1, wherein the protection system action is a warning. 4. The system of claim 1, wherein the protection system action is an alarm. 5. The system of claim 1, wherein the protection system action includes a reactor scram. 6. The system of claim 1, wherein the protection system action includes a warning. 7. The system of claim 1, wherein the protection system action includes an alarm. 8. The system of claim 1, wherein the device configured to initiate a protection system action initiates a reactor scram when the monitored reactor power is greater than the power-dependent high reactor pressure setpoint that corresponds to a value of percent power in the operating domain of the reactor. 9. The system of claim 1, wherein the device configured to initiate a protection system action initiates a reactor scram when the monitored reactor power is greater than the power-dependent high reactor pressure setpoint that corresponds to 100% reactor power. 10. The system of claim 1, further comprising:one or more signals in the system that correspond to one or more values of percent reactor power;wherein the one or more signals in the system that correspond to one or more values, respectively, of percent reactor power are delayed in time before the one or more signals in the system that correspond to one or more values, respectively, of percent reactor power affect the power-dependent high reactor pressure setpoint. 11. The system of claim 1, further comprising:one or more signals in the system that correspond to one or more values of percent reactor power; andone or more signals in the system that correspond to reactor pressure;wherein the one or more signals in the system that correspond to one or more values, respectively, of percent reactor power are lagged relative to the one or more signals, respectively, in the system that correspond to reactor pressure. |
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abstract | A nuclear fuel assembly having a plurality of multi-leaf hold down spring sets extending from a top nozzle. Each spring set consists of a multiple number of springs leafs in order to provide a large working range of spring deflection. Each spring leaf has a straight, flat base section followed by a straight, flat tapered beam with a secondary spring set having a curvature at its peripheral end. |
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description | This application is a continuation of and claims priority under 35 U.S.C. §120 to application Ser. No. 12/078,705 filed Apr. 3, 2008 now U.S. Pat. No. 7,970,095, the entirety of which is incorporated by reference. 1. Field Example embodiments generally relate to fuel structures and radioisotopes produced therein in nuclear power plants. 2. Description of Related Art Generally, nuclear power plants include a reactor core having fuel arranged therein to produce power by nuclear fission. A common design in U.S. nuclear power plants is to arrange fuel in a plurality of fuel rods bound together as a fuel assembly, or fuel assembly, placed within the reactor core. These fuel rods typically include several elements joining the fuel rods to assembly components at various axial locations throughout the assembly. As shown in FIG. 1, a conventional fuel assembly 10 of a nuclear reactor, such as a BWR, may include an outer channel 12 surrounding an upper tie plate 14 and a lower tie plate 16. A plurality of full-length fuel rods 18 and/or part length fuel rods 19 may be arranged in a matrix within the fuel assembly 10 and pass through a plurality of spacers 20. Fuel rods 18 and 19 generally originate and terminate at upper and lower tie plates 14 and 16, continuously running the length of the fuel assembly 10, with the exception of part length rods 19, which all terminate at a lower vertical position from the full length rods 18. An upper end plug 15 and/or lower end plug 17 may join the fuel rods 18 and 19 to the upper and lower tie plates 14 and 16, with only the lower end plug 17 being used in the case of part length rods 19. Tie rods 28 may be full length rods placed at corner positions in fuel assembly 10 that securely join to upper and lower tie plates 14 and 16 and provide handling points for fuel assembly 10. The end plugs 15 and 17 may mate with, and in the case of tie rods 28, pass through, the upper and lower tie plates 14 and 16, respectively, and may secure fuel rods 18 or 19 axially in the fuel assembly 10. Example embodiments are directed to tie plate attachments having irradiation targets and fuel assemblies that use example embodiment tie plate attachments and methods of using the same to generate radioisotopes. Example embodiment tie plate attachments may include a plurality of retention bores that permit irradiation targets to be inserted and contained in the retention bores. The irradiation targets may be irradiated in an operating nuclear core including the fuel assemblies, generating useful radioisotopes that may be harvested from the spent nuclear fuel assembly by removing example embodiment tie plate attachments. Example embodiment tie plate attachments may be connected to fuel assemblies via the upper tie plate, fuel rods, and/or channel surrounding the fuel assembly. Example embodiment tie plates may be held at a fixed axial position within fuel assemblies so as to expose irradiation targets therein to constant, lower-level neutron flux, thereby converting a substantial amount of the irradiation targets into useable radioisotopes. Detailed illustrative embodiments of example embodiments are disclosed herein. However, specific structural and functional details disclosed herein are merely representative for purposes of describing example embodiments. The example embodiments may, however, be embodied in many alternate forms and should not be construed as limited to only example embodiments set forth herein. It will be understood that, although the terms first, second, etc. may be used herein to describe various elements, these elements should not be limited by these terms. These terms are only used to distinguish one element from another. For example, a first element could be termed a second element, and, similarly, a second element could be termed a first element, without departing from the scope of example embodiments. As used herein, the term “and/or” includes any and all combinations of one or more of the associated listed items. It will be understood that when an element is referred to as being “connected,” “coupled,” “mated,” “attached,” or “fixed” to another element, it can be directly connected or coupled to the other element or intervening elements may be present. In contrast, when an element is referred to as being “directly connected” or “directly coupled” to another element, there are no intervening elements present. Other words used to describe the relationship between elements should be interpreted in a like fashion (e.g., “between” versus “directly between”, “adjacent” versus “directly adjacent”, etc.). The terminology used herein is for the purpose of describing particular embodiments only and is not intended to be limiting of example embodiments. As used herein, the singular forms “a”, “an” and “the” are intended to include the plural forms as well, unless the language explicitly indicates otherwise. It will be further understood that the terms “comprises,” “comprising,” “includes,” and/or “including,” when used herein, specify the presence of stated features, integers, steps, operations, elements, and/or components, but do not preclude the presence or addition of one or more other features, integers, steps, operations, elements, components, and/or groups thereof. It should also be noted that in some alternative implementations, the functions/acts noted may occur out of the order noted in the figures. For example, two figures shown in succession may in fact be executed substantially concurrently or may sometimes be executed in the reverse order, depending upon the functionality/acts involved. FIG. 2 illustrates an example embodiment fuel assembly 100 including upper tie plate 114 and an example embodiment tie plate attachment 150 that individually or together may function as a radioisotope production structure. Example embodiment fuel assembly 100 may be similar to conventional fuel assemblies with the exception of including example embodiment tie plate attachment 150. Although example embodiment fuel assembly 100 is shown as similar to a conventional BWR type fuel assembly, other example embodiments, including PWR type fuel assemblies and unfinished fuel bundles, may be useable with tie plate attachments according to the present invention. Example embodiment tie plate attachment 150 may be generally rectangular and frame full-length fuel rods 118 in fuel assembly 100. An outer perimeter of example embodiment tie plate attachment 150 may extend to about an outer perimeter of fuel assembly 100 formed by fuel rods 118 so as to form a substantially uniform axial profile within example embodiment fuel assembly 100. Although example embodiment tie plate attachment 150 is shown as generally rectangular with a hollow center, other shapes are possible. For example, example embodiment tie plate attachments may extend along only one or two sides of example embodiment fuel assemblies instead of all four sides. Similarly, example embodiment tie plate attachments may have varied thicknesses or even extend through the entire cross-sectional profile of example embodiment fuel assemblies and have channels permitting coolant flow therethrough instead of having a hollow center. Example embodiment tie plate attachments may also have other shapes to match example embodiment fuel assemblies and tie plates therein, including hexagonal, triangular, etc. shapes. In the example embodiment tie plate of FIG. 2, attachment 150 may have a cross-sectional edge thickness equal to a single row of fuel rods 118 along a transverse cross section of example embodiment fuel assembly 100. That is, example embodiment tie plate attachment 150 may surround, or be co-located with, the outer fuel rods 118 in example embodiment assembly 100. In this way, example embodiment tie plate attachment 150 may not significantly reduce or interfere with coolant flow through interior rods in assembly 100 and may be placed at a position with typically lower neutron flux within the assembly 100. As shown in FIG. 2, example embodiment tie plate attachment 150 may be positioned under upper tie plate 114 in an axial direction. Example embodiment tie plate attachment 150 may be held under upper tie plate 114 in a variety of ways. For example, example embodiment attachment 150 may be directly welded to upper tie plate 114, forged into or be otherwise structurally continuous with upper tie plate 114, may fit into upper tie plate 114 frictionally and/or in a lock-and-key fashion, or may be joined to upper tie plate 114 via fasteners such as bolts or screws. As shown in FIGS. 2 and 3, as another attachment option, example embodiment tie plate attachment 150 may permit one or more fuel rods 118 and/or upper end and tie plugs 120 to pass axially through attachment 150 via holes 155 and into upper tie plate 114. Fuel rods 118 may thus fix example embodiment tie plate attachment 150 in a transverse position under upper tie plate 114. Example embodiment tie plate attachment 150 may be held in a constant axial position under tie plate 114 by fuel rods 118 seating into holes 155 or by flow of coolant through assembly 100 in an axial direction, and/or fixing example embodiment tie plate attachment 150 against upper tie plate 114. Or, for example, fuel rods 118 and/or upper end plugs 120 may be screwed into, locked into, welded onto, etc., example embodiment tie plate attachment 150 so as to hold attachment 150 in a constant axial position under upper tie plate 114. Even further, example embodiment tie plate attachment 150 may attach to outer channel 112 by being welded and/or removably fitted into outer channel 112 surrounding example embodiment fuel assembly 100. Lateral extensions (discussed below) may facilitate such contact between outer channel 112 and example embodiment tie plate attachment 150. In example embodiment fuel assemblies, example embodiment tie plate attachments may thus be held near or attached under an upper tie plate in the axial direction. This position affords easy access to example embodiment tie plate attachments during assembly disassembly, as the example embodiment tie plate attachment may be accessed with removal of the upper tie plate alone. FIG. 3 is a detailed illustration of an example embodiment tie plate attachment 150. Although example embodiment tie plate attachment 150 is shown as a hollow rectangle that matches the shape of the outer channel 112, other shapes and orientations are possible as discussed above. Example embodiment tie plate attachment 150 is fabricated of a material that substantially maintains its physical and neutronic properties when exposed to conditions in an operating nuclear core, such that example embodiment tie plate attachment 150 does not interfere with or affect the neutron flux present in the operating reactor. Example embodiment tie plate attachments may be fabricated of, for example, stainless steel, Inconel, a nickel alloy, a zirconium alloy, aluminum, etc. As discussed above, holes 155 may penetrate entirely through example embodiment tie plate attachment 150 and permit fuel rods 118 (shown in shadow) and/or upper end plugs 120 to pass through and/or connect to example embodiment tie plate attachment 150. As such, holes 155 may be sized with an inner diameter sufficiently greater than a fuel rod 118 and/or upper end plug 120 outer diameter. The example joining method of FIG. 3 shows example embodiment tie plate attachment 150 “sitting” on the shoulder 117 of the fuel rod 118 and upper end plug 120 joint. It is understood and several other joining methods discussed above and below may be used, including frictional contact between rods or end plugs and example embodiment tie plate attachments, lock-and-key, slot-type, or dovetail-type joints, welding, and/or continuous connection between the parts. Example embodiment tie plate attachment 150 may include one or more lateral extensions 165 that facilitate positioning relative to and/or connection with channel 112. For example, lateral extensions 165 may connect or abut channel 112 on each side of example embodiment tie plate attachment 150 in order to center and/or secure example embodiment tie plate attachment 150 within example embodiment fuel assembly 100. Lateral extensions 165 may further match extensions and/or shape of the upper tie plate 114 in order to provide a consistent axial profile among upper tie plate 114 and example embodiment tie plate attachment 150. Example embodiment tie plate attachment 150 includes a plurality of retaining bores 160 in its top face into which one or more irradiation targets 170 are placed and contained, as shown in FIG. 4, which is a blown up portion of area A in FIG. 3. Bores 160 do not pass through example embodiment tie plate attachment 150 but instead have a depth sufficient to allow irradiation targets 170 to fit within bores 160. Bores 160 may be geometrically placed around or between holes 155. Alternatively, bores 160 may be scattered in no particular pattern throughout example embodiment tie plate attachment 150, so long as the structural integrity of attachment 150 is not compromised by the position and/or number of bores 160. Irradiation targets 170 may be in the shape of small “seeds” or small rod shapes for insertion into retaining bores 160. Based on the size of bores 160, irradiation targets 170 may have a width and length to fit within bores 160 and may be, for example, on the scale of millimeters. Several irradiation targets 170 containing potentially different types of parent materials, including solids, liquids, and/or gasses, may be placed into a single retaining bore 160. Alternatively, each bore 160 may contain homogenous irradiation targets 170. Irradiation targets 170 may be made of a variety of materials that substantially convert into radioisotopes when exposed to a neutron flux encountered under tie plates 114 in an operating nuclear reactor. Because neutron flux may be lower at axial ends of example embodiment fuel assembly 100 (FIG. 2), example embodiment tie plate attachments and irradiation targets 170 therein may be exposed to a lower flux as well. Hence, materials having high neutron cross sections and shorter half-lives may be preferable for use as irradiation targets 170, including, for example, Iridium-191, which may convert to Iridium-192 when exposed to neutron flux encountered in an operating nuclear reactor. Similarly, other isotopes, including Cobalt-59, Selenium-74, Strontium-88, and/or Iridium-191 for example, may be used as irradiation targets 170. Retention bores 160 may be sealed or closed by a cap 161, shown in FIG. 4, that covers bores 160 and joins to example embodiment tie plate attachment 150. For example, caps 161 may be welded onto attachment 150 or screwed into bores 160, if the bores 160 are threaded. Other methods of securely attaching caps 161 over bores 160 in order to provide containment of irradiation targets 170 may be known and useable with example embodiments. Because cap 161 may provide containment to retention bores 160, irradiation targets 170 may contain or produce useful gaseous, liquid, and/or solid radioisotopes when exposed to a neutron flux, and these radioisotopes may be contained in irradiation bores 160 by cap 161 even though they may be liquid, gaseous, or solid. Because of the higher axial position of example embodiment tie plate attachments, irradiation targets contained therein may be irradiated by lower amounts of neutron flux over a longer period of time, resulting in more predictable and effective generation of radioisotopes with shorter half-lives from irradiation targets having higher cross sections. Further, because upper tie plate areas, where example embodiment tie plate attachments may be placed, are associated with low fretting, example embodiment tie plate attachments may provide robust containment for irradiation targets. Lastly, upper tie plates may be easily removed from irradiated example embodiment fuel assemblies without disturbing fuel rods or irradiated fuel, permitting easier harvesting of example embodiment tie plate attachments and useful radioisotopes therein. Example embodiment tie plate attachments may further provide robust containment for retaining and containing solid, liquid, or gas radioisotopes produced from irradiation targets in example embodiment tie plate attachments. Example embodiments thus being described, it will be appreciated by one skilled in the art that example embodiments may be varied through routine experimentation and without further inventive activity. For example, other fuel types, shapes, and configurations may be used in conjunction with example embodiment fuel assemblies and tie plate attachments. Variations are not to be regarded as departure from the spirit and scope of the exemplary embodiments, and all such modifications as would be obvious to one skilled in the art are intended to be included within the scope of the following claims. |
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abstract | The present disclosure is directed towards methods and systems and methods for measuring the integrity of an operating system's execution and ensuring that the system's code is performing its intended functionality. This includes examining the integrity of the code that the operating system is executing as well as the data that the operating system accesses. Integrity violations can be detected in the dynamic portions of the code being executed. |
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abstract | Steam turbine plant includes a steam generator, a plurality of low pressure turbines being driven by steam from the steam generator, a plurality of steam condensers to condense the steam from the low pressure turbines into condensed water and a feedwater line which supplies the condensed water to the steam generator as feedwater. The feedwater line including a plurality of feedwater heating lines connected in parallel. A number of feedwater heating lines being less than a number of steam condensers. Each of the feedwater heating lines includes at least one low pressure feedwater heater provided in at least one of the steam condensers to heat the condensed water by steam bled from the low pressure turbines. |
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abstract | The invention is intended to shorten a positioning time required for forming an irradiation area with high accuracy using a number of leaf plates, and to reduce physical and mental burdens imposed on patients. A multi-leaf collimator comprises leaf plate driving body each including a plurality of movable leaf plates and provided respectively on one side and the other side, the plurality of leaf plates of the leaf plate driver on one side and the plurality of leaf plates of the leaf plate driver on the other side being disposed in an opposing relation to form an irradiation field of a radiation beam between the opposing leaf plates. Each of the leaf plate driving body includes a motor provided in common to the plurality of leaf plates. Driving force of the motor can be transmitted to the plurality of leaf plates at the same time through a pinion gear, upper and lower air cylinders, and upper and lower guides. Also, the driving force can be cut off selectively for each leaf plate. |
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049851837 | summary | BACKGROUND OF THE INVENTION The present invention relates to a process for fabricating nuclear reactor fuel pellets having large grain sizes from highly active UO.sub.2 powder. In particular, the present invention concerns a method for controlling the sintered density of UO.sub.2 pellets to a predetermined range. When UO.sub.2 pellets are used as the fuel in nuclear reactors, it is important that the fuel density be high within a reasonable limit so that a more compact reactor core can be designed, and that the thermal conductivity of the pellets is sufficiently high. However, when the sintered density of the pellets is too high, swelling of the pellets during irradiation becomes too great, thereby damaging a tube in the reactor. Accordingly, UO.sub.2 pellets commonly used in light-water reactors are usually designed so that the sintered density is in the range of from 94 to 97% TD (theoretical density). One recent technical innovation is to prolong the useful life of the reactor fuel. This is called the plan for "high burnup", and it is now being studied seriously. In order to execute this plan, it is imperative to restrain the fission gas (FP gas) in the pellets as much as possible. It is well known that producing large crystal grain sizes is effective in confining FP gas in the pellets. However, the conventional technology only produced grain sizes of at most about 10 to 20 .mu.m. In light of the above, the applicants have proposed a process for fabricating UO.sub.2 pellets with large-grain size crystals in JP-A-62-297215, JP-A-63-45127, U.S. patent application Ser. Nos. 139447, 296802 and 296808. These processes have a common effect of producing in that they make crystalline grains of large size by controlling the conditions of ammonium diuranate (ADU) formation. With the processes described in the applications and patents, it is possible to control the crystalline grain size, however at the time these previous patents were filed, the applicants did not consider to control the sintered density of the pellets. In other words, when pellets are fabricated pellets having grain sizes larger than 20 .mu.m by the process described above, the sintered density of the resulting pellets are as high as 98 to 99% TD. In order to reduce the sintered density of the sintered body, if required, it has been common to add a pore-former agent to the raw material powder, which cause the formation of pores when the agent sublimates during sintering. The applicants believe that the method is applicable to the fabrication process for pellets composed of large-grain size crystals. Although a pore-former agent of this kind is effective in reducing the sintered density, the agent is likely to have adverse effects on UO.sub.2 grain growth. In other words, the formation of crystalline grains with large grain size is disrupted. Accordingly, it is desired to develop a method for controlling both the sintered density of the pellets and the crystalline grain sizes. SUMMARY OF THE INVENTION Accordingly, it is an object of the present invention to provide a process for fabricating UO.sub.2 pellets having the sintered density of pellets within the range of from 94 to 97% TD, when pellets having large crystalline grain sizes in excess of 20 .mu.m are fabricated from highly active UO.sub.2 powder, and thereby restrain the fission gasses generated during the irradiation in the pellets thereof to enhance the safety of irradiation. According to an aspect of the present invention, there is provided a process for fabricating UO.sub.2 pellets comprising (1) uniformly adding at least one pore-former agent in the range of 0.3 to 1.4% by weight, to uranium dioxide powders to form a starting material, the pore-former agent decomposing and sublimating below 600.degree. C. and having average grain size in the range of 5 to 500 .mu.m, (2) compacting the uranium dioxide powders, including the pore-former agent, to form green pellets, and (3) sintering the green pellets to form sintered UO.sub.2 pellets having large grain-size crystals. According to the present invention, Accordingly, both the sintered density of the pellets and the crystalline grain sizes can be easily controlled to the desired range, which was not previously possible. DETAILED DESCRIPTION A process for fabricating UO.sub.2 pellets, according to the present invention, will be specifically described below. First, a pore-former agent is added to highly active UO.sub.2 powder. From experimental results, the inventors have determined that the following conditions must be met in order to maintain a sintered density in the range between 94 to 97% TD. (1) It is necessary that the particle size of the pore-former agent be between 5 and 500 .mu.m, and preferably between 10 and 100 .mu.m. When the particle size is less than 5 .mu.m, the pores left behind after the sublimation of the agent during sintering disrupted the growth of the crystals, and thus, the crystalline grain sizes in the pellets is small. On the other hand, when the particle size exceeds 500 .mu.m, large pores are formed on the surface of pellets. These large pores must be avoided because they accelerate the absorption of the water into the pellets. (2) It is necessary to use a pore-former agent that decomposes and sublimates below 600.degree. C., more preferably at 500.degree. C. When the decomposition temperature is higher than 600.degree. C., the pore-former agent is confined to the inside of the pellets during the sintering process and cracks or holes appear in the pellets. Ammonium acetate, ammonium carbonate, ammonium bicarbonate, ammonium oxalate, ammonium alginate, stearic acid, and the like, are pore-former agents that meet the above requirements. These compounds are used either alone or in a mixture. (3) The appropriate amount of these pore-former agent is in the range of 0.3 to 1.4% by weight of the UO.sub.2 powders. The sintered density of the pellet will not be in the range of 94 to 97% TD, when the added amount falls outside this range. Subsequently, UO.sub.2 powder, to which a pore-former agent has been added, is fabricated into pellets by any of the methods described below. (a) UO.sub.2 powder, to which a pore-former agent has been added, is filled into a mold and pressed to make a pressed body. (b) A lubricant is added uniformly to UO.sub.2 powder to which a pore-former agent has been added. Thus prepared, the powder is filled into a mold and subjected to compacting. (c) UO.sub.2 powder, to which a pore-former agent has been added, is filled into a mold coated with a lubricant. Subsequently, the compacting step is carried out on the filled mold. (d) UO.sub.2 powder that includes a pore-former agent is roughly molded into a lump which is pulverized to obtain granules. The granules are compacted by any of the processes described in (a) to (c) above. Stearic acid, zinc stearate, lithium stearate, stearic amide, ethylene-bis-stearic amide, methylene-bis-stearic amide, polyethylene glycol, and the like, are suitable compounds for the lubricant. These compounds are used either alone or in mixtures. When a lubricant of this kind is added to the raw material powder, or when a lubricant is applied to the mold, the press-compacting is easily carried out. When the granules are made by the method described in (d), the size of the granules should be less than 2000 .mu.m, and more preferably less than 1000 .mu.m. When the size exceeds 2000 .mu.m, it is difficult to fill a designated amount of the granule into a mold and perform the compacting step. |
044951370 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS FIG. 1A and 1B are schematic views showing an embodiment of the present invention. The outside of a reactor vessel 1 is surrounded by a guard vessel 2 with a predetermined distance therebetween, and a liquid manometer structure 3a is mounted on the guard vessel 2. The space S between the reactor vessel 1 and the guard vessel 2 is communicated with the external space only through the manometer structure 3a. When a sealing liquid 4 fills the manoeter structure 3a, the space S inside the guard vessel 2 is perfectly sealed. Reference numeral 5 represents a reactor core. In this embodiment, a substantially U-shaped pipe (generally, a large diameter pipe of stainless steel), which has a sufficient inner diameter to permit welders and welding tools to pass through, is employed for the liquid manometer structure 3a. To produce the arrangement of this embodiment in practice, the guard vessel 2 is first welded where necessary to the reactor vessel 1 from both outside and inside, and the U-shaped pipe or the manometer structure 3a, is then reliably welded to the upper part of the outer surface of the guard vessel 2. After the welding is finished, the welders and the welding tools can leave through the inside of the U-shaped pipe 3a, which is subsequently filled with the sealing liquid 4. In this manner, a highly reliably sealed space S can be formed between the reactor vessel 1 and the guard vessel 2. Even when the sealed space S is pressurized, the U-shaped pipe 3a establishes a balance between the pressures inside and outside of the space due to the principle of a liquid manometer. It is therefore necessary to decide upon the length of a rising pipe 9 outside the U-shaped pipe 3a in consideration of the rise of the liquid level. Metals which remain in a liquid state under the conditions of use, such as low-temperature melting alloys, may be used as the sealing liquid 4, but liquid metals having a high specific gravity such as mercury can also be used advantageously. To bring the space S between the reactor vessel and the guard vessel into a pressurized state, inert gases of the like can be charged into the space. Incidentally, in the case of a loop type reactor, the piping in the primary cooling system close to the reactor vessel main body, that is, a coolant outlet pipe 6 and a coolant inlet pipe 7, are also surrounded by a guard pipe 8 connected to the guard vessel 2. In the present invention, the space between these coolant pipes 6 and 7 and the guard pipe 8 is also sealed. Since these coolant pipes 6 and 7 and the guard pipe 8 can be extended to the outside of a biological shield W, however, the space between these coolant pipes and the guard pipe can be sealed by using conventionally known bellows 10 outside the shield W. Since the bellows portion 10 is disposed outside the shield W, it can be replaced whenever necessary. As the diameter of each of the coolant pipes 6 and 7 is far smaller than the diameter of the reactor vessel, the space between the pipes 6 and 7 and the guard pipe 8 could be liquid-sealed relatively easily using a small annular liquid manometer structure 11 as shown in FIG. 2. When such a small but reliable annular manometer structure 11 is employed, the space between the coolant pipes and the guard pipe can be sealed inside the shield W, as shown in FIG. 2. Next, FIGS. 3A and 3B and FIGS. 4A and 4B show other embodiments of the present invention, respectively. In these embodiments, the liquid manometer structure 3b and 3c are disposed around the entire circumference of the upper portions of the reactor vessel 1 and guard vessel 2. In these embodiments, since the construction other than the liquid manometer structure is the same as that of the embodiment shown in FIGS. 1A and 1B, like reference numerals are used to identify like constituent elements and their explanation is omitted. In the embodiment shown in FIGS. 3A and 3B, an annular flange 12 is formed extending outward from the upper peripheral edge of the guard vessel 2 in the horizontal direction, and the tip of this flange portion 12 is shaped into a groove 13 having a substantially U-shaped cross section. An annular member 14 having an inverted L-shaped cross section or a sideways T-shaped cross section is also formed extending outward from the outer circumferential surface of the reactor vessel 1 in the horizontal direction above the extending flange portion 12. The lower end portion 14a of the annular member 14 is inserted into the U-shaped groove 13 so that both members 12 and 14 are coupled with each other and a gap is maintained between the lower end portion 14a and the U-shaped groove 13. According to this arrangement, the liquid manometer structure 3b can be formed around the entire circumference at a position separated from the reactor vessel 1. In contrast, FIGS. 4A and 4B show an embodiment in which the liquid manometer sturcture 3c is formed close to the reactor vessel 1. An annular member 15 having an L-shaped cross section is disposed on the inner circumference at the upper portion of the guard vessel 2 so as to define a groove structure. On the other side, an annular member 16 having an inverted L-shaped cross section or a sideways T-shaped cross section is disposed around the outer circumference of the reactor vessel 1, and these members 15 and 16 are fit together so as to leave a gap between the lower end 16a of the member 16 and the member 15, thereby forming the liquid manometer structure 3c. Besides the abovementioned liquid manometer structures, various other structures may also be used in the present invention. For instance, the liquid manometer sturcture 3e in the embodiment shown in FIG. 5 consists of an annular member 17 having an L-shaped cross section disposed around the outer circumference at the upper portion of the guard vessel 2 so as to define a groove structure, and an annular member 18 having an inverted L-shaped cross section or a sideways T-shaped cross section disposed around the outer circumference of the reactor vessel 1, both members 17 and 18 being combined with each other to form the liquid manometer structure 3e. According to the construction shown in FIGS 3A, 4A and 5, no welded portion exists between the reactor vessel 1 and the guard vessel 2, so the sealed space S that is formed therebetween can have especially high reliability. In the nuclear reactor having such a construction, the reactor vessel 1 and the guard vessel 2 are combined with each other only via the liquid-sealed portion, so that no stress transfered between the two vessels, which is especially desirable for the reactor arrangement. When the nuclear reactor is a fast breeder, the temperature is likely to exceed 300.degree. C. in the proximity of the reactor vessel 1 and guard vessel 2 at the cover gas region in FIGS. 4A and 5. When exposed to such a temperature, the liquid which fills in the manometer structure of the present invention would likely evaporate. In order to prevent the evaporation of the sealing liquid, heat insulating materials or cooling means may be additionally provided lest the heat of the reactor vessel directly affect the liquid manometer structure. When the sealed space S defined between the reactor vessel 1 and the guard vessel 2 is brought into the pressurized state, the pressure may be kept substantially equal to the pressure inside the reactor vessel. However, the pressure need not always be equal to the internal pressure of the reactor vessel. In a pressurized state of about 1.5 to about 3 atms of the absolute pressure, leakage of coolant from damaged portions of the reactor vessel and the piping in the primary cooling system can be markedly reduced. Accordingly, even if a coolant leakage occurs, the flow rate of the coolant inside the core can be kept substantially equal to the flow rate during normal operation of the reactor and damage to the core can be prevented. Thus, it becomes possible to prevent possible accidents. In addition, when the pressure change inside the sealed and pressurized space S is monitored by using a pressure gauge and the like, the occurrence of an accident can be detected. If a tag gas is charged into the sealed space S defined between the reactor vessel 1 and the guard vessel 2 to bring the space S into a pressurized state, the occurrence of an accident can be quickly detected by monitoring leakage of the tag gas from a crack when such is generated in the reactor vessel, in the guard vessel or in the piping in the primary coolant system. FIG. 5 shows a preferred embodiment for monitoring leakage of the tag gas. Namely, the space S between the reactor vessel 1 and the guard vessel is sealed by the liquid manometer structure 3e and the tag gas fills the space S. A pressure gauge 19 is disposed in order to measure the internal pressure of the sealed space S. A level meter 20 is immersed in the liquid 4 of the manometer structure 3e. Further, tag gas detection means 22 and 23 are positioned in the atmospheric region in an upper cover gas space 21 within the reactor vessel 1 and outside the guard vessel 2, respectively. Since the rest of the construction is substantially the same as that of the embodiment shown in FIG. 1, like reference numerals are used to identify like constituents and the explanation of these members are omitted. When damage such as a crack occurs in the reactor vessel 1 or in the pipes 6 and 7 in the primary cooling system connected to the reactor vessel, the tag gas charged in the space S comes into the reactor vessel 1 from the damaged portion and leaks into the cover gas space 21 at the upper part of the reactor vessel or into a cover gas space (not shown) in an instrument installed in the primary cooling system. The tag gas is immediately detected by the tag gas detection means, e.g. means 22 disposed inside the cover gas space 21, so that the damage of the reactor vessel 1 can be detected at an early stage. If damage develops on the guard vessel 2, the tag gas leaks to the external region and is likewise detected by the detection means 23 for the tag gas. The leakage of the tag gas can be detected most conveniently by charging such a tag gas at a predetermined pressure within the range of, for example, about 1.5 to about 3 atms (absolute pressure) into the sealed space S, and by measuring the pressure change inside the space S by the pressure gauge 19 mounted on the guard vessel 2. The pressure change inside the space S can also be supervised by measuring the level change of the liquid 4 of the manometer structure 3e by the level meter 20. In measuring the pressure change, it is necessary to compensate for the volume change of the tag gas due to any temperature change inside the space S. On the basis of the pressure measurement by the pressure gauge 19 and level meter 20 and the result of tag gas leak detection by the detection means 22, 23, it becomes possible to always determine accurately whether or not damage to the reactor vessel or the guard vessel has occured. As can be understood from the foregoing, either one of the pressure gauge 19 and level meter 20 may be provided, or both may be disposed conjointly. It is not always necessary to provide both pressure gauge 10 or level meter 20 and detection means 22, 23 for the tag gas. Either one may be provided. The tag gas to be charged into the sealed space S is preferably an inert gas such as helium, other rare gases, or mixtures thereof. Any means may be employed as the detection means 22, 23 for the tag gas employed, so long as they are capable of detecting the presence of the gas, but it is generally preferred to employ a gas sampling device and a mass spectrometer associated with the device. In this case, the tag gas preferably contains such a component that can be sharply sensed by the mass spectrometer. The leakage of the tag gas can be supervised rapidly and reliably if a rare gas containing its specific stable isotope or a gas mixture containing isotopes in a specific mixing ratio is used as the tag gas. Examples of the stable isotopes of the rare gases include neon 20, neon 21 and neon 22. These examples are merely illustrative and not restrictive in any manner. In contrast to the abovementioned construction, it is also possible to keep the sealed space S between the guard vessel and the reactor vessel under negative pressure and to dispose the gas detection means inside the sealed space S so that a crack in the guard vessel is detected by detecting the leakage of gas from the external region into the sealed space S. As described in detail in the foregoing, in the nuclear reactor in accordance with the present invention, the space between the reactor vessel and the guard vessel is kept in a reliably sealed and pressurized state. Accordingly, reactor damage in the cases of a core disassembly accident, that can be assumed theoretically, and breakage accidents on the reactor vessel or the piping in the primary cooling system can be largely reduced, and the safely of the reactor can be markedly improved. In addition, by filling the tag gas in the sealed space to bring the space into a pressurized state and by detecting the leakage of the tag gas from the sealed space, the occurence of accident, or damage to the reactor vessel, the guard vessel, the piping in the primary cooling system and the like can always be quickly detected even during reactor operation. While the invention has been described with respect to preferred embodiments, it should be apparent to those skilled in the art that numerous modifications may be made thereto without departing from the spirit and scope of the invention. |
claims | 1. An electromagnetic wave interference (EMI)/radio frequency interference (RFI) shielding resin composite material, comprising:(A) a thermoplastic polymer resin;(B) about 20 to about 69 volume % of an electrically conductive filler having a polyhedral shape or being capable of forming a polyhedral shape;(C) a low-melting point metal including a primary component comprising tin, bismuth, lead, or a combination thereof, and a secondary component comprising copper, aluminum, nickel, silver, germanium, indium, zinc, or a combination thereof; and(D) a glass fiber filler. 2. The EMI/RFI shielding resin composite material of claim 1, wherein the EMI/RFI shielding resin composite material comprises:about 30 to about 85 volume % of the thermoplastic polymer resin (A);about 1 to about 10 volume % of the low-melting point metal (C). 3. The EMI/RFI shielding resin composite material of claim 1, comprising the glass fiber filler (D) in an amount of about 50 parts by weight or less based on about 100 parts by weight of the EMI/RFI shielding resin composite material. 4. The EMI/RFI shielding resin composite material of claim 1, wherein the thermoplastic polymer resin (A) comprises a polyamide, a polyalkylene terephthalate, a polyacetal, a polycarbonate, a polyimide, a polyphenylene oxide, a polysulfone, a polyphenylene sulfide, a polyamide imide, a polyether sulfone, a liquid crystal polymer, a polyetherketone, a polyetherimide, a polyolefin, acrylonitrile-butadiene-styrene, a polystyrene, a syndiotactic polystyrene, or a combination or blend thereof. 5. The EMI/RFI shielding resin composite material of claim 1, wherein the electrically conductive filler having a polyhedral shape or being capable of forming a polyhedral shape (B) comprises a needle-shaped electrically conductive filler having a polyhedral interior, a sheet-shaped electrically conductive filler having a polyhedral interior, a globular electrically conductive filler having a polyhedral interior, or a combination thereof. 6. The EMI/RFI shielding resin composite material of claim 5, wherein the needle-shaped electrically conductive filler having a polyhedral interior is a metal filler fabricated in a needle shape by pressing and cutting a dendrite metal filler fabricated through an electrolysis process or a porous metal filler fabricated through a thermal process, or a needle-shaped metal filler fabricated by polishing a metal lump;the sheet-shaped electrically conductive filler having a polyhedral interior is a metal filler fabricated in a sheet shape by pressing a dendrite metal filler fabricated through an electrolysis process or a porous metal filler fabricated through a thermal process, or a sheet-shaped metal filler fabricated through a pulverization process; andthe globular electrically conductive filler having a polyhedral interior is a globular metal filler fabricated through a melt injection process. 7. The EMI/RFI shielding resin composite material of claim 1, wherein the electrically conductive filler (B) is broken down or pulverized by a shear stress applied during a process of making the EMI/RFI shielding resin composite material to thereby form a polyhedral shape. 8. The EMI/RFI shielding resin composite material of claim 7, wherein the electrically conductive filler (B) has a shear strength of under about 300 MPa. 9. The EMI/RFI shielding resin composite material of claim 1, wherein the electrically conductive filler (B) comprises aluminum, copper, magnesium, iron, nickel, molybdenum, zinc, silver, alloys thereof or a combination thereof. 10. The EMI/RFI shielding resin composite material of claim 1, wherein the low-melting point metal (C) is a solid solution including two or more kinds of metal elements. 11. The EMI/RFI shielding resin composite material of claim 1, wherein the low-melting point metal (C) has a solidus temperature that is lower than a temperature of a process of making the EMI/RFI shielding resin composite material. 12. A molded product made using the EMI/RFI shielding resin composite material according to claim 1. 13. An electromagnetic wave interference (EMI)/radio frequency interference (RFI) shielding resin composite material, comprising:(A) a thermoplastic polymer resin;(B) about 20 to about 69 volume % of an electrically conductive filler having a polyhedral shape or being capable of forming a polyhedral shape;(C) a low-melting point metal having a solidus temperature that is lower than a temperature of a process of making the EMI/RFI shielding resin composite material; and(D) a glass fiber filler. |
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041815718 | summary | BACKGROUND OF THE INVENTION This invention relates generally to nuclear reactors but more particularly to bracing grids for nuclear reactor fuel sub-assemblies. A nuclear reactor may have a fuel assembly comprising a plurality of fuel sub-assemblies each sub-assembly comprising a bundle of spaced fuel pins enclosed within a tubular wrapper. Where the fuel pins are closely spaced and slender as in a sub-assembly for a liquid metal cooled fast breeder nuclear reactor they may be spaced apart by a longitudinal series of bracing grids disposed at intervals along the sub-assembly to ensure that the coolant subchannels bounded by the fuel pins are maintained continuous. A typical bracing grid comprises a plurality of strips of material formed to produce a complex of hexagonally shaped cells in honeycomb pattern. Each sub-assembly may contain 250 to 350 fuel pins and the bracing grids are required to be manufactured to very close tolerances to ensure that the pins can be inserted in the longitudinal series of grids during assembly. A typical nuclear reactor of this kind with honeycomb bracing grids is disclosed in United States Pat. No. 4,036,690. Adverse tolerances in the construction of the grid can result in severe damage to the pins on insertion during assembly and on withdrawal during disassembly when the irradiated pins are fragile and it is an object of the invention to provide for a nuclear reactor a fuel sub-assembly having fuel pin bracing grids possessing a degree of compliancy whereby the fuel pins are enabled to be inserted and withdrawn collectively of the grids without causing serious damage. SUMMARY OF THE INVENTION According to the invention in a fuel pin bracing grid for a nuclear fuel sub-assembly and comprising a plurality of strips formed to produce a complex of hexagonally shaped unit cells in honeycomb pattern, each cell is formed from a discrete strip of material, each of three alternate sides of the cell having a window or opening formed therein and the remaining sides being continuous, each continuous side having an embossment to provide a guide pad for a fuel pin. The windows reduce the stiffness of the regions intermediate the ends of the cells thereby endowing the grid with sufficient compliancy for, according to one aspect, the pins to deflect resiliently the cells of successive grids on entry with low insertion loads and without causing serious damage by scoring of the pins, and, according to a second aspect, to enable the pins to be withdrawn by low withdrawal loads thereby minimising damage to brittle irradiated pins. The unit cell construction provides a strong grid with compliancy, it facilitates manufacture by jigs which maintain accurate the geometrical relationship of the cells and simplifies manufacture because the cells can be joined together by edge welds applied collectively from each face of the grid. In a preferred construction of bracing grid according to the invention the continuous sides of each unit cell have a pair of second embossments, the second embossments of each pair being disposed one on each side of the first embossments to form a linear group extending parallel to the longitudinal axis of the cells, the second embossments being of lesser radial height than the first embossments. The preferred construction of bracing grid has two levels of compliancy, firstly the compliancy of the intermediate regions of the cells as afforded by the windows and secondly the lesser compliancy afforded by the stiffer end regions of the cells wherein all the sides of the end regions are uninterrupted. The second embossments can provide additional supports for the fuel pins to resist bowing during irradiation and thereby facilitate withdrawal of the bundle of fuel pins from the wrapper and grid. The invention will reside in a nuclear fuel sub-assembly comprising a central fuel section extended at opposite ends by a locating section and a neutron shielding section and wherein the central fuel section comprises a multiplicity of elongate fuel pins spaced apart within a tubular wrapper by a series of bracing grids, each grid comprising a plurality of strips formed to produce a complex of hexagonally shaped unit cells in honeycomb pattern, each cell being formed from a discrete strip of material, each of three alternate sides of the cell having a window or opening formed therein and the remaining sides being continuous, each continuous side having an embossment to provide a guide pad for a fuel pin. A fuel sub-assembly embodying the invention has the advantage that it can be readily dismantled by cutting the wrapper to separate the end sections from the fuel section and withdrawing the fuel pins collectively from the wrapper and grid combination thereby greatly facilitating reprocessing. The invention will also reside in a liquid metal cooled fast breeder nuclear reactor of the kind wherein a nuclear fuel assembly is submerged in a pool of coolant, the fuel assembly comprising a plurality of fuel sub-assemblies upstanding in side-by-side array, each sub-assembly comprising a multiplicity of elongate fuel pins spaced apart within a tubular wrapper by a series of bracing grids each grid comprising a plurality of strips formed to produce a complex of hexagonally shaped unit cells in honeycomb pattern, each cell being formed from a discrete strip of material, each of three alternate sides of the cell having a window or opening formed therein and the remaining sides being continuous, each continuous side having a first embossment to provide a guide pad for a fuel pin and a pair of second embossments disposed one on each side of the first embossment to form a linear group extending parallel to the longitudinal axis of the cells, the second embossments being of lesser radial height than the first embossments. A liquid metal cooled fast breeder nuclear reactor of the pool kind has a fuel assembly comprising a multiplicity of slender fuel pins which are susceptible to severe bowing under irradiation; the second embossments of the bracing grids provide additional lateral support for the pins to restrain that bowing whenever the supports provided by the first embossments are inadequate because of the compliant nature of the grid. |
039873064 | description | DETAILED DESCRIPTION In FIG. 1 a U.V. radiation generating apparatus comprises a chamber 10 formed by metal end plates 11 sealed to the respective open ends of a tubular metal portion 12 shaped such that the chamber has a uniform elliptical transverse cross-section. The inner surfaces of the chamber are polished such that they are highly reflective in the U.V. region of electromagnetic radiation. An electrode assembly 13 is sealingly mounted in each end plate 11 with the longitudinal axes of the assemblies 13 substantially coincident with the straight line locus of one of the foci of the elliptical section of the chamber. A hollow tube 14 formed of quartz passes through the chamber substantially coaxial with the locus of the other focus, and is sealingly secured to the end plates by means of nuts 15 and associated O-rings 16. The hollow tube 14 defines a passage through the chamber for liquid to be treated by U.V. radiation. The liquid to be treated is supplied via pump 107 and flow rate control means 108. Inlet and outlet tubes 17 are mounted in the wall of the tubular portion 12 for enabling a flow of liquid through the chamber. The discharge region 18 between the two electrode assemblies 13 is initially completely occupied by the liquid and upon application of a high voltage between the electrode assemblies 13 an electrical discharge ocurs therebetween. Liquid in and adjacent the discharge region vaporizes and tries to expand but this expansion is resisted by the inertia of the liquid. If there are pumping means associated with the inlet and outlet tubes for pumping the liquid through the chamber, the liquid may have no free outlet and be effectively confined to the chamber and the expansion is further resisted by the incompressible nature of the liquid. The liquid vapour in the discharge region attains an extremely high temperature and pressure resulting, by the choice of an appropriate liquid, in the emission of U.V. light. Because of the constrained discharge region this method of generating U.V. light is more efficient than a conventional method employing a gas discharge lamp. The two electrode assemblies 13 are identical and each comprises a solid cylindrical stainless steel electrode 19 having an integral enlarged head portion 20 of substantially conical form with a radiussed tip portion. The electrode 19 is threaded along its length and is screwed into an insulating sleeve 21 which has its end adjacent the enlarged head 20 shaped to the same conical section as the head 20 so that there is no discontinuity of the surface of the assembly at the junction between the sleeve 21 and the enlarged head 20. The sleeve is formed from a phenolic laminate. The sleeve 21 is sealingly mounted in its respective end plate by means of an enlarged diameter flange portion 22 at the outer end of the sleeve, an externally threaded portion on the sleeve mating with an internally threaded mounting hole in the end plate, and an O-ring seated within a channel in the mating surface of the flange portion 22. A lock nut 23 and washer 24 are provided on the electrode 19 in order to prevent relative movement between the electrode and its sleeve. Alternative forms of the above described apparatus of FIG. 1 may differ in one or more of the following respects: the shape of the enlarged head portion 20 may be hemispherical instead of conical; the electrode 19 may be threaded only in the region where locknut 23 is to be mounted, and the sleeve 21 may not be internally threaded, or may only be threaded over a portion of its bore; the enlarged head portion 20 may be non-integral and secured to the electrode by, for example, a screw thread or soldering; the metal of the electrode may be tungsten; thorium oxide may be added to the stainless steel or tungsten to increase the electron emission; the tip region of the head portion may be formed as an insert of a material different from that of the electrode and head portion; one or both of the electrode assemblies may have means, for example a separate locknut on the externally threaded portion of the sleeve, for enabling the position of the electrode assembly to be adjusted; the transverse cross-section of the tubular portion 12 may be circular instead of elliptical; either or both of the electrode assemblies may be mounted in the wall of the tubular portion; there may be a plurality of hollow tubes 14 mounted parallel to, and closely spaced from, each other; the insulating sleeves may be formed from another resilient material for example polyethylene or polytetrafluoroethylene; the tubular portion and the end plates may be formed from a plastics material, or a glass or ceramic material; if required, the material of the end plates may be different to that of the tubular portion; the inner surfaces of the chamber may be made reflective by means of metallising or electroplating with a suitable metal, for example, chromium. The positions of the electrode assemblies may be adjusted to alter the dimensions of the discharge region 18 upon initially setting-up the discharge conditions and subsequently in order to compensate for erosion of one or both of the head portions of the electrode assemblies. An arrangement for enabling adjustment of the position of an electrode is shown schematically in FIG. 11 where a sealing device 102 around sleeve 21 seals to end plate 11 and to sleeve 21. A driving means 103 is responsive to the electrical conditions of the discharge and is arranged to turn sleeve 21 in dependence thereupon. The diameter of the bore of the hollow tube(s) 14 is chosen to obtain a desired rate of flow of liquid to be treated by U.V. radiation. The maximum linear velocity of liquid in the tube(s) is determined by the repetition rate of the electrical discharge, in other words, the maximum velocity is that beyond which there is less than 100% irradiation of the liquid flowing through the hollow tube(s). The wall thickness of the hollow tube(s) is chosen such that the tube has sufficient mechanical strength to withstand pressure fluctuations generated by the electrical discharges. Quartz is a suitable material since it provides the necessary mechanical strength and also is substantially tansparent in the U.V. region of the radiation emitted from the discharge region. Where the transverse cross-section of the tubular portion is circular, the electrode assemblies may be disposed longitudinally along the axis of the tubular portion; and where there is a plurality of hollow tubes 14, these may be parallel to this axis and arranged symmetrically therearound at a convenient radius. The electrical circuit shown in FIG. 2 comprises a high voltage D.C. power supply 27 connected to a capacitor 28 via limiting resistor 29. In practice, resistor 29 may be mostly formed by the secondary resistance of the high voltage tansformer utilised in the power supply 27. At a suitable point in the charging cycle, normally at or near the maximum voltage attainable on the capacitor, a high-speed high voltage switch 30 is actuated by means (not shown) to connect the capacitor to the electrode assemblies 13 in the chamber 10, thus causing an electrical discharge in the liquid in the chamber. The high voltage D.C. supply 27 provides an unsmoothed unidirectional output, using either full or half-wave rectification. The means for actuating the switch 30 is synchronised with the frequency of the mains source for the supply 27 to provide a repetitive electrical discharge. Thus provided that capacitor 28 can be sufficiently charged in half a period of the mains frequency, the electrical discharges can be maintained at a frequency of twice the mains frequency. The discharge circuit parameters (i.e. the resistance and inductance of the discharge current path from capacitor 28 via switch 30 to electrode assemblies 13; and also the value of the capacitor 28, and the voltage applied across it) are all chosen in such a way as to result in the discharge occuring in chamber 10 having the required characteristics, as regards pressure, temperature and duration. Typical values might be: ______________________________________ capacitance 0.2.mu.F voltage 20 kV inductance 2 .mu.H resistance 0.5 .OMEGA. ______________________________________ resulting in typical discharge parameters as: ______________________________________ temperature 15,000.degree.K pressure 10 Kbar duration 20 .mu.sec ______________________________________ But any other values may be incorporated in the circuit, as may seem appropriate to adapt the discharge of different types of U.V. treatment. The liquid within which the electrical discharge is to be effected should be at least partially, and preferably substantially, transparent to U.V. radiation. The electrical characteristics of the liquid should be such as to give an appropriate form of discharge, between the electrodes. A convenient liquid is water, either distilled or industrial. As an alternative to introducing and removing the water via inlet and outlet tubes 17, either or both of the electrode assemblies may be adapted to provide a liquid flow channel for delivery or extraction of water. Thus the chamber may have a single tube 17, either or both of the electrode assemblies having a flow channel; the direction of flow may be in either direction i.e. into or out of the tube 17. In another arrangement the chamber may have no tube 17 and in this case both of the electrode assemblies will have flow channels, the water being introduced via one of the electrode assemblies and removed via the other assembly. FIG. 3 shows one form of electrode assembly having a channel for the introduction or removal of the water. The electrode 19 has a cylindrical bore 31. At the inner end of the electrode the bore communicates with the conical surface of the head portion via a plurality of symmetrically arranged short passages 32. The axes of the passages 32 are arranged approximately normal to the conical surface of the head portion. In FIG. 4, which shows a modification of FIG. 3, the electrode 19 is provided with a cylindrical bore which passes centrally through the head portion and issues at tip of the head portion. FIG. 5 shows a different form of electrode assembly having such a channel. In this assembly, a solid threaded cylindrical electrode portion 33 is supported within a large bore in an electrode 19. An internally threaded plug 34 supports the electrode portion 33 centrally and seals thereto and to the outer end of electrode 19. An apertured spacer 35 disposed near the inner end of the assembly maintains the correct spacing of the electrode portion 33 within the electrode 19. A passage 36 communicates through the wall of electrode 19 with the annular flow channel between electrode portion 33 and electrode 19. The position of the electrode portion 33 can be adjusted by means of the plug 34 in order to alter inter-electrode gap. If required the sleeve may also have an adjusting arrangement as mentioned in connection with the apparatus of FIG. 1. FIG. 6 shows a modification of the assembly of FIG. 5 in which the electrode 19 is omitted, and the flange 22 of the sleeve 21 is adapted to support directly an electrode portion 33' similar to portion 33. An inlet passage 37 is provided in the flange 22 and communicates with the annular flow channel 109 around electrode portion 33'. In the electrode assembly shown in FIG. 7, the electrode is in the form of a hollow cylindrical tube 38, a portion of the outer surface of which is threaded to engage internal threads at the flange of the sleeve 21. The tube 38 is maintained in position by means of locknut 23 and washer 24. The outer end of tube 38 is adapted to be connected to a hose or pipe for the introduction or removal of water, and the inner end protrudes a short way beond the frusto-conical surface of the inner end of sleeve 21. The annular inner end surface of tube 38 is substantially flat and normal to the axis of the tube. FIG. 8 shows another embodiment of an apparatus for generating U.V. radiation. The chamber of the apparatus comprises a cup-shaped lower portion 40 formed of stainless steel and having an external flange 41 around its rim. Sealed to the rim if portion 40 by means of O-ring 42 is a disc of quartz 43, and an end plate 44 formed of stainless stel is sealed to the upper surface of disc 43 by means of O-ring 45. End plate 44 is held secure by means of a plurality of securing-bolts passing through the peripheral regions of the end plate 44 and the flange 41. The underside of end plate 44 has a circular recess 46 surrounded by an annular channel 47 which leads to an outlet passage 48 in the peripheral region of end plate 44. The depth of the recess 46 is made approximately equal to the maximum depth of penetration of U.V. radiation of appropriate wavelength in the liquid medium to be treated. A central inlet passage 49 introduces liquid to be treated through the end plate 44 into the narrow circular chamber provided by recess 46 in conjunction with the upper surface of disc 43. This liquid flows radially outwardly, substantially in the form of a thin film, into the channel 47 and then flows out via passage 48. The inner surfaces of portion 40 and end plate 44 are rendered optically reflective by polishing, although other methods of achieving this may be used as previously mentioned. An electrode assembly 13, substantially identical to those in the apparatus of FIG. 1, is axially mounted in the bottom of the portion 40. An inlet pipe 50 introduced water into the interior of portion 40 close to its bottom and is positioned such that the incoming flow is substantially tangential to the transverse cross-section shape of the portion 40 at the level of inlet pipe 50. An outlet pipe 51 is set into the side wall of portion 40 a short distance (typically 5 mm) above the level of the tip of the electrode assembly and defines the static surface level of water in the chamber. The portion 40 acts as an electrode, electrical connection being made thereto by means of a threaded hole 52. The effect of introducing water into the chamber as a tangential flow results in a circulating motion of the water in the chamber. This motion causes a depression of the water surface 53 above the electrode assembly and a corresponding rise of the surface near the walls of portion 40. The amount of this depression is dependent upon the rotational speed of the water and can thus be altered by controlling the inlet flow by a pump 104 and fluid control valve 105 linked to the high voltage generating circuit, in known manner, automatically to maintain the electrical parameters of successive discharges at predetermined values. In this way the effect of wear, erosion or other loss of material from the tip of the electrode may be compensated for. The electrical discharge is from the tip of the electrode to approximately the nearest point of the water surface 53 and is completed by "tracking" or further electrical breakdown across the water surface. Since the "tracking" will occur along random radial paths to the side walls of portion 40, there will be a spreading of the effects of wear and erosion. Also, since as a result of wear of the electrode assembly the rotational speed of the water is increased, the level of the discharge termination points will vary and further spread the effects of wear. The height of portion 40 is selected such that disc 43 will not be damaged by droplets or jets of water thrown up from the discharge region. The portion 40 has its inner surface shaped such as to direct radiation from the discharge region in the direction of the disc 43. Such a shape may be, for example, a paraboloid of revolution arranged such that the discharge region is approximately at the focus of the paraboloid. Alternatively, the lower part of the inner surface may be hemispherically arranged such that the discharge region is approximately at the centre of the hemisphere. The flow rate of liquid to be treated is fed via pump 107 (FIG. 1) and flow control means 108 (FIG. 1) which is controlled in dependence upon the discharge rate such that there is at least 100% irradiation. If required, the time taken for liquid to flow from inlet passage 49 into channel 47 may be made several times the period of the electrical discharges. FIG. 9 shows another embodiment of U.V. radiation generating apparatus comprising substantially identical upper and lower stainless steel chamber portions 60 and 61, respectively, which are clamped together by securing bolts (not shown) passing through respective flanges 62, 63, and sealed together via O-ring 64. The chamber has a cylindrical side wall 65 and flat upper and lower end walls 66, 67. Mounted in the lower chamber portion 61 axially with respect to the chamber is an electrode assembly substantially as utilised in the apparatus of FIG. 1. An electrode assembly as shown in FIG. 6 is correspondingly mounted in the upper chamber portion 60. Water introduced into the chamber via inlet passage 37 of the upper electrode assembly flows downwardly through the annular passage of this assembly and exits at the open tip of the sleeve of the assembly. The falling column of water crosses the discharge region between the electrode assemblies and then flows downwardly over and around the lower electrode assembly, continuously enveloping the tip of head portion of the lower electrode assembly. A drain passage 68 is provided in the lower chamber portion 61 for the removal of water accumulating in the bottom of the chamber. The two electrode assemblies are arranged such that the discharge region between them is substantially at the middle of the chamber cavity, the electrical discharge being substantially along the axis of the column of water. The diameter of the bore in the sleeve of the upper electrode assembly determines the radius of the column of water at the discharge region, and is chosen to provide the required temperature and pressure of the plasma generated in the discharge region. A thin-walled cylinder of quartz 69 is secured betwen annular grooves 70, 71 in the inner surfaces of end walls 66, 67, respectively. The wall thickness of about 1 mm. O-rings 72 space the cylinder 69 from the inner cylindrical surface of side walls 65 to form a treatment cavity 73 in the shape of a cylindrical annulus; the O-rings also seal the treatment cavity 73 from the cavity of the chamber. The thickness of the cavity 73 will be determined by the maximum penetration of U.V. radiation in the liquid to be treated. The radius of the quartz cylinder 69 is approximately 20 mm and is chosen such that there should be no damage caused by water droplets driven out of the discharge region by the explosive force of the electrical discharge, and also such as to obtain a desired intensity of U.V. radiation at the treatment cavity. An inlet passage 74 is provided in the bottom of the cylindrical wall 65 for introducing liquid to be treated into the cavity 73, and an outlet passage 75 is provided in the top of the wall 65 diametrically opposite passage 74 for the outflow of the liquid from cavity 73. The rate of flow of liquid to be treated and the electrical discharge repetition rate are interdependent as mentioned previously. If a discharge rate at the frequency of the mains supply, or some multiple thereof, is not suitable, than an alternative electrical circuit may comprise an electronic pulse generator of known form adapted to operate at any required frequency and coupled to a known form of high-voltage triggered spark gap. If it is required to prevent a particular wavelength or small range of wavelengths of the generated radiation from entering the treatment cavity 73, the outer surface of the quartz cylinder 69 may have a thin film of absorbing material deposited thereon. Alternatively, a thin film may be deposited thereon of such thickness as to prevent transmission by means of optical interference. Alternatively, or additionally, the quartz may be "doped" with a suitable metallic ion, for example, tungsten. A protective layer of magnesium fluoride would be deposited on top of any thin film on the outer surface of the quartz cylinder 69 in order to avoid any chemical interaction between such thin film and the liquid to be treated. The inner metal surfaces are polished so as to be highly reflective at U.V. wavelengths. Alternatively, other methods may be used, as mentioned previously. The inner cylindrical surface of side wall 65 may have thin films for selective absorption deposited thereon in the same manner as the outer surface of quartz cylinder 69. Alternatively, or additionally, a thin film may be deposited thereon for the selective reflection of specific wavelengths or a range of wavelengths. In an alternative form of the apparatus of FIG. 9, the upper electrode assembly may be of the form shown in FIG. 7, in which case the radius of the bore of tube 38 will determine the radius of the column of water at the discharge region. The electrical discharge will occur from a point on or near to the inner edge of the annular end face of tube 38, downwardly along the surface of the water column to a point near the head portion of the lower electrode assembly, and thence through the body of the water column to a point on the surface of the head portion of the electrode assembly. The radius of the bore of tube 38 will be chosen to provide a required submerged-discharge path length which results in a discharge of required characteristics. FIG. 10 shows another embodiment of an apparatus for generating U.V. radiation which comprises a lower chamber portion 80 formed of stainless steel and having a flat bottom wall 81 and a cylindrical inner side wall 82. An annular ring portion 83 formed of insulating material is mounted on top of lower chamber portion 80 by means of securing bolts 84 (not shown) in conjunction with flange 85 on chamber portion 80, and O-ring 86. Mounted on top of ring portion 83 by means of bolts 84 and O-rings 87 and 88 is a quartz disc 43 and an end plate 44' which differs from the end plate 44 used in the apparatus FIG. 8 only in that the outlet passage 48 is parallel to the inlet passage 49 instead of being at right angles thereto. Electrode carriers 89 and 90 are mounted diametrically opposite each other in ring portion 83. Electrode carrier 89 comprises a threaded rod having an enlarged diameter unthreaded head portion 91. Electrode 92 formed from 5 mm diameter stainless steel bar has one end secured in an angled bore passing through head portion 91. A set screw 93 engaging in a threaded axial bore communicating with the angled bore serves to hold the electrode 92 secure. The depth of penetration of the electrode carrier 89 into the chamber cavity, and the effective length of electrode are chosen such that the working end face of 94 of electrode 92 is substantially on the axis of the chamber. Electrode carrier 90 is similar to electrode carrier 89 but is arranged to carry a 1 mm diameter stainless steel electrode 95 having a working end face 96, and is arranged such that the end face 96 is adjacent end face 94 at substantially the same distance from bottom wall 81. Near the bottom of chamber portion 80 there is an inlet passage 97 connected to a supply (not shown) of mercury having an adjustable head such that the height, nominally about 1 cm, of mercury in the bottom of the chamber cavity can be altered. The surface 98 of the mercury is nominally about 5 mm below the end face 94 of electrode 92. An inlet passage 99 communicates with the chamber cavity for the introduction of water at approximately the same level as the level of end face 94. Inlet passage 99 is arranged such that the incoming water flow is approximately tangential to the inner cylindrical surface of the chamber. This is so that a circulating motion can be transferred to the mercury for the purpose of maintaining the characteristics of electrical discharges between end face 94 and surface 98 of the mercury which acts as an electrode, the chamber portion 80 being electrically connected to a pole of an electrical discharge circuit 106 (a modified form of the circuit of FIG. 2) by means of screw 100 engaged in a threaded bore in the chamber portion 80. The inlet flow of water from pump 104 is controlled by fluid control valve 105 as mentioned previously. The water is removed from the chamber cavity via an outlet passage 101 in flange 85. Electrode carriers 89 and 90 are electrically connected to the electrical discharge circuit 106 which comprises a portion arranged to apply a voltage of about 1kV between electrode 95 and the chamber portion 80 to establish a degree of ionisation of the mercury in the vicinity of the discharge region between electrode 92 and the mercury surface 98. This discharge is called an initiating discharge. This technique is well-known and is used for example in a device called an ignitron. The energy of such initiating electrical discharge is of the order of 1 Joule. Discharge circuit 106 comprises another portion arranged to apply a voltage of about 20kV between electrode 92 and the chamber portion 80 in synchronism (by the use of a known discharge circuit) with the initiating discharge to cause an electrical discharge of the required characteristics. The energy of this latter discharge is in the range 10 to 100 Joules. The radiation emitted from the plasma of mercury vapour in the discharge region is at a wavelength of 254 nm and is particularly appropriate to the killing or deactivation of several strains of bacteria, yeast spores and fungi. The initiating discharge can, if required, be effected between the electrode 95 and electrode 92. Whereas in the apparatus of FIG. 10 the chamber portion 80 is formed of stainless steel, any other convenient material may be used, and, if necessary, a protective coating may be applied to the inner surfaces by for example chromium plating or vacuum deposition. The inner surfaces may have additional coatings for selective reflection or absorption of radiation at selected wavelengths, the coatings be covered by a protective layer of for example magnesium fluoride as mentioned previously. If the protective or other coatings are electrically insulating, electrical connection to the mercury may be effective by omitting these coatings over, say, the bottom wall 81 (if chamber portion 80 is conductive). Where the material of chamber portion 80 is non-conductive, electrical connection may be made to the mercury by using, a conductive screw 100 and arranging for the threaded bore to communicate with the chamber cavity. It will be appreciated that other conductive liquids or solutions may be used instead of the mercury, (the lower liquid), and other liquids instead of the water (the upper liquid). The choice of such liquids will depend on the desired wavelengths of the emitted radiation and on the electrical characteristics of the liquids, also, the upper liquid must be less dense than, and immiscible with, the lower liquid. The material of the chamber portion 80, or as the case may be, the protective layer or coating, is selected to have substantial immunity to chemical or electrochemical attack by either of the liquids in the chamber cavity, or their vapours, either separately, or in any combination. If the ring portion 83 is electrically conductive at the high voltages it is subjected to, or if it made from a normally conductive material, the electrode carriers 89 and 90 may be mounted by means of insulating bushings (not shown). The electrode 92 and 95 may be formed of tungsten or some other suitable conductive material. If necessary, the surface of the electrodes, apart from their working ends, may be covered with a protective coating as mentioned previously. If it is not required to cause rotation of the mercury for depressing the surface immediately below the electrode 92, 95 then it will not be necessary to have a flow of water through passages 99 and 101, in other words the water in the chamber cavity can be substantially static. In the aforedescribed embodiments the electrical connection to the electrodes is conveniently made by providing a tag or similar electrical connecting terminal (not shown) under the screw 23 at the outer end of the electrode assembly. It will be appreciated that either of a pair of electrodes may be connected to the more positive pole of the electrical discharge circuit. It will be appreciated that the various alternative forms of electrode assemblies may be interchanged in the apparatus shown as required and as appropriate. |
abstract | A patient alignment system for a radiation therapy system. The alignment system includes multiple external measurement devices which obtain position measurements of components of the radiation therapy system which are movable and/or are subject to flex or other positional variations. The alignment system employs the external measurements to provide corrective positioning feedback to more precisely register the patient and align them with a radiation beam. The alignment system can be provided as an integral part of a radiation therapy system or can be added as an upgrade to existing radiation therapy systems. |
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description | 1. Field of the Invention The present invention relates to garments, and more particularly, to garments for protection from ultraviolet radiation. 2. Description of the Related Art Skin cancer is a cancer that starts in the skin. Some other types of cancer start in other parts of the body and can spread to the skin, but these are not skin cancers. There are two main types of skin cancers, keratinocyte cancers, and melanomas. Basal and squamous cell skin cancers are by far the most common cancers of the skin. They start in cells called keratinocytes, the most common cells in the skin. Melanomas are cancers that develop from melanocytes. The cells that make the brown pigment that gives skin its color. Melanocytes can also form benign growths called moles. There are other types of skin cancers as well, but they are much less common. They include merkel cell carcinoma, kaposi sarcoma, cutaneous lymphoma, skin adnexal tumors, and various types of sarcomas. However, together, these types account for less than 1% of all skin cancers. Exposure to ultraviolet (UV) radiation is a major risk factor for most skin cancers. Sunlight is the main source of UV rays. People who get a lot of UV exposure from the sun are at greater risk for skin cancer. Even though UV rays make up only a very small portion of the sun's rays, they are the main cause of the sun's damaging effects on the skin. UV rays damage the DNA of skin cells. Skin cancers start when this damage affects the DNA of genes that control skin cell growth. There are three main types of UV rays. UVA rays age skin cells and can damage their DNA. These rays are linked to long-term skin damage such as wrinkles, but they are also thought to play a role in some skin cancers. UVB rays can directly damage skin cells' DNA, and are the main rays that cause sunburns. They are also thought to cause most skin cancers. Lastly, UVC rays don't get through our atmosphere and are not in sunlight. They are not normally a cause of skin cancer. Both UVA and UVB rays damage skin and cause skin cancer. UVB rays are a more potent cause of at least some skin cancers, but based on what's known today, there are no safe UV rays. The amount of UV exposure a person gets depends on the strength of the rays, the length of time the skin is exposed, and whether the skin is protected with clothing or sunscreen. Skin cancers are one result of getting too much sun, but there are other effects as well. Sunburn and tanning are the short-term results of too much exposure to UV rays, and are signs of skin damage. Long-term exposure can cause early skin aging, wrinkles, loss of skin elasticity, dark patches, and pre-cancerous skin changes. Applicant believes that one of the closest references corresponds to U.S. Patent Application Publication No. 20120255094 A1, published on Oct. 11, 2012 to Victor Dragony for sun screen article. However, it differs from the present invention because Dragony teaches a sun screen article protecting at least an arm and a shoulder, and optionally a portion of the neck and/or a portion of the hand of a person wearing it from excessive exposure to solar radiation, and method of using the sun screen article inside a vehicle. The article includes a tubular portion adapted to protect an arm and a flap adjoining the tubular portion and adapted to protect a shoulder. An optional collar portion affixed to the open and of the flap is structured to screen at least a portion of the neck. An epaulette, in affixable cooperation with the flap of the sun screen article, prevents a shoulder harness from freely moving with respect to the sun screen article and a body of the sun screen article from sliding down the arm of the wearer. Applicant believes that another reference corresponds to U.S. Patent Application Publication No. US 20100024088 A1, published on Feb. 4, 2010 to Shannon Griefer for UV protected arm sleeves. However, it differs from the present invention because Griefer teaches an arm sleeve comprising an upper arm portion, a lower arm portion, and a pocket, wherein the arm sleeve provides protection against harmful ultraviolet rays. The arm sleeve may be made out of spandex or a combination of spandex and any one or more of bamboo, polyester, nylon, hemp, maize, lyocell, or other wood pulp based fabric, or other synthetic or natural knitted or woven fabric. The arm sleeve may also have a fastener to attach two or more arm sleeves together. Applicant believes that another reference corresponds to U.S. Pat. No. 6,775,844 B1 issued to Patrick Castillo on Aug. 17, 2004 for arm shades. However, it differs from the present invention because Castillo teaches a health apparatus for use by individuals while driving. The health apparatus would be an arm shield, which would be worn over an individual's “outside arm” while driving. The arm shield would reduce sun exposure on the arm and would be attached via two end-mounted elastic bands. An extra hood could be wrapped around the individual's hand on the outside arm for added protection. Applicant believes that another reference corresponds to U.S. Pat. No. 6,539,550 B1 issued to Barbara Flores on Apr. 1, 2003 for a set of driving gloves. However, it differs from the present invention because Flores teaches a set of three driving gloves, each having a different length to be worn by a driver. Each glove has a varying length with finger portions cut away to allow for greater flexibility when driving. The set includes a full length glove, a medium length glove, and a short glove. Applicant believes that another reference corresponds to U.S. Pat. No. 6,029,278 A issued to Guillermo Lopez on Feb. 29, 2000 for a sun protection device. However, it differs from the present invention because Lopez teaches a device for protecting the user while seated in a vehicle from the harmful effects of the sun. The sun protection device includes a headpiece, which may be a cap or headband, a face/neck cover for shielding the side of the user's face and directly exposed to the sun and a shoulder/arm cover for shielding the user's shoulder and arm directly exposed to sunlight. The device may further include a hand cover for the user's hand, which is most directly exposed to sunlight, as well as a second shoulder/arm cover for the side of the user's body indirectly exposed to the sun through other vehicle openings. Applicant believes that another reference corresponds to U.S. Pat. No. 5,628,062 A issued to Li Ming Tseng on May 13, 1997 for an arm and hand UV protection sleeve for driving. However, it differs from the present invention because Tseng teaches an arm and hand ultra violet protection sleeve for driving that includes a special graded fabric sun-block sleeve for UV protection while driving. The UV-proof sleeve is constructed of special graded soft and smooth irritation free fabric material, and with an elongated air ventilating chamber-like cavity extended from the upper arm portion down to the wrist area and from the wrist, a cuff extends in arch over the back of the hand which ends over the tip of the fingers, and with fastening elements and openings to both ends, so that the sleeve can be held in place gently and worn comfortably while driving. The UV protection sleeve can effectively narrow down and reduce the chance of drivers contracting any type of skin damage or health hazardous skin diseases from the intrusion of ultra violet radiation. Applicant believes that another reference corresponds to U.S. Pat. No. 5,357,633 A issued to George V. Rael on Oct. 25, 1994 for an arm protective garment. However, it differs from the present invention because Rael teaches an arm protective garment that includes an elongated tubular sleeve made of a flexible fabric and defining an elongated internal cavity extending between opposite ends. The sleeve is open at one end for slipping over a driver's hand and arm and for receiving the driver's arm in the internal cavity of the sleeve. The garment also includes a mitten of flexible fabric disposed on the other end of the sleeve and defining an internal pocket for receiving the driver's hand therein. The mitten has only a thumb opening defined therein for extension of the driver's thumb from the mitten. The garment further includes a flexible strap attached to the one open end of the sleeve for encircling the neck or chest of the driver for releasable reattachment to the one open end of the sleeve to retain the sleeve on the driver's arm. Applicant believes that another reference corresponds to U.S. Pat. No. 5,056,157 A issued to Linda D. Pryor on Oct. 15, 1991 for a solar radiation protecting device and method. However, it differs from the present invention because Pryor teaches a solar radiation protecting device for protecting the forearm and perhaps a portion of an upper arm of an individual when the arm of that individual projects outwardly of a vehicle window while the passenger or driver is seated within the vehicle. This protective device will protect against excessive solar radiation exposure when the arm is so projected beyond the window of the vehicle and thereupon exposed to solar radiation. The protecting device comprises a flexible fabric covering which extends over at least a portion of the forearm of the individual and also permits air exposure to the skin while worn. Strap means are located on the flexible fabric covering for releasably securing the fabric covering to the forearm of the individual. Applicant believes that another reference corresponds to U.S. Pat. No. D675,381 S issued to Patricia Rambo on Jan. 29, 2013 for a sun protective garment. However, it differs from the present invention because Rambo teaches a different design from that of Applicant. Applicant believes that another reference corresponds to U.S. Pat. No. D649,293 S issued to Frank T. Lyons on Nov. 22, 2011 for an arm protector for blocking sunlight while driving. However, it differs from the present invention because Lyons teaches a different design from that of Applicant. Applicant believes that another reference corresponds to EP Patent No. 1754420 (A2) issued to Draznin Elke on Feb. 21, 2007 for a sun protective sleeve for car driver. However, it differs from the present invention because Elke teaches a sleeve made of a light textile material for a comfortable feeling on a hot day and is designed as an ordinary sleeve provided with an extension covering the back of the hand. The upper end can also be provided with an extension in order to prevent the sleeve from exposing the shoulder of the user. The sleeve is available in various sizes. Applicant believes that another reference corresponds to KR Patent No. 100819426 B1 issued to Sin Dong II on Apr. 4, 2008 for a sun cover for arm. However, it differs from the present invention because Sin Dong II teaches an arm cover for protecting an arm from the sunlight to achieve a smooth air circulation and a hygienic usage, and to allow a user to have a convenient wearing feeling. An arm cover for protecting an arm from a sunlight includes: an upper-arm covering unit, which is to be fastened to an upper arm of a user; and a forearm covering unit, which is to be fastened to his/her forearm. The upper-arm covering unit comprises: a first upper-arm covering member, having a shape corresponding to his/her upper arm; and a second upper-arm covering member, wrapping an outer surface of the first upper-arm covering member; as well as an air vent. The forearm covering unit comprises: a first forearm covering member, having a shape corresponding to his/her forearm; a second forearm covering member, wrapping an outer surface of the first forearm covering member; and a hand back covering member, having a film member, and connected to the first forearm covering member in a swiveled manner. Other patents describing the closest subject matter provide for a number of more or less complicated features that fail to solve the problem in an efficient and economical way. None of these patents suggest the novel features of the present invention. The instant invention is a garment for protection from ultraviolet radiation, comprising a torso garment having a front side and a rear side that extend from a first edge to an end. The torso garment further comprises first and second lateral sides and first and second shoulder sections. Extending from the first and second lateral sides and the first and second shoulder sections are first and second sleeves respectively. First and second hand covers extend from the first and second sleeves respectively. The first and second hand covers each comprise an elastic band. The elastic band, a distal end, and third and fourth lateral sides define an interior face. The interior face comprises a thumb loop and at least first and second finger loops. The elastic band, the distal end, and the third and fourth lateral sides also define an exterior face. The interior and exterior faces may fold internally within the first and second sleeves respectively, or may fold externally onto the first and second sleeves respectively. In a preferred embodiment, the elastic bands are sewn to the first and second hand covers. The thumb loop is positioned at a predetermined distance from the elastic band without reaching the distal end. The torso garment further comprises securing means to secure a neck gaiter. The torso garment further comprises a neckband that extends from the first edge. The neckband may also comprise securing means to secure the neck gaiter. The securing means includes fasteners, spring snaps assemblies, hook and loop fasteners, and zipper assemblies. The securing means can be positioned at an interior or exterior side of the neckband. The spring snaps assemblies comprise caps, sockets, studs, and posts. The neck gaiter comprises an exterior side and an interior side that extend between a top end and a bottom end. The neck gaiter also comprises securing means to secure onto the neckband. The torso garment, first and second hand covers, and neck gaiter are made of stretchable fabrics/materials such as spandex, cotton, cotton blends, nylon, polyesters, and combinations thereof. It is therefore one of the main objects of the present invention to provide a garment for protection from ultraviolet radiation that is easily worn on the upper torso of a user. It is another object of this invention to provide a garment for protection from ultraviolet radiation that covers the back of a user's hands. It is another object of this invention to provide a garment for protection from ultraviolet radiation that is versatile. It is another object of this invention to provide a garment for protection from ultraviolet radiation that is volumetrically efficient. It is another object of this invention to provide a garment for protection from ultraviolet radiation that is comfortable. It is another object of this invention to provide a garment for protection from ultraviolet radiation that is durable. It is yet another object of this invention to provide such a device that is inexpensive to manufacture and maintain while retaining its effectiveness. Further objects of the invention will be brought out in the following part of the specification, wherein detailed description is for the purpose of fully disclosing the invention without placing limitations thereon. Referring now to the drawings, the present invention is a garment for protection from ultraviolet radiation and is generally referred to with numeral 10. It can be observed that it basically includes torso garment 20, hand covers 40, and neck gaiter 70. As seen in FIGS. 1 and 2, torso garment 20 comprises front side 22 and rear side 24 that extend from edge 36 to end 28. Torso garment 20 further comprises lateral sides 30 and shoulder sections 26. Extending from lateral sides 30 and shoulder sections 26 are respective sleeves 32. Extending from each sleeve 32 is a respective hand cover 40. Each hand cover 40 comprises elastic band 42. In a preferred embodiment, elastic band 42 snugly and comfortably fits at, or approximately at, a user's wrist section, whereby elastic band 42 is sewn thereon. Elastic band 42, distal end 56, and lateral sides 50 and 52 define interior face 48. Sewn onto interior face 48 is thumb loop 54. Thumb loop 54 is designed to receive a thumb of a user. In a preferred embodiment, thumb loop 54 is positioned at a first predetermined distance from elastic band 42 without reaching distal end 56. Also sewn onto interior face 48 are at least first and second finger loops 58 and 60. As seen for illustrative purposes, first and second finger loops 58 and 60 are designed to receive a right index and little finger respectively of the user. It is understood that the other illustrated hand cover 40 is designed for a left hand of the user. Elastic band 42, distal end 56, and lateral sides 50 and 52 also define exterior face 62 that is designed cover a back of the user's hand. In a preferred embodiment, present invention 10 is worn to fit the user so that only the user's fingers extend beyond distal end 56, and lateral side 50 being adjacent to the user's thumb. In an alternate embodiment, each interior face 48 and/or exterior face 62 has semi-ridged properties so that it may fold internally within sleeve 32, or externally onto sleeve 32, at elastic band 42 to remain in place in the event the user wants to cover his/her arms but not the back of the hands. Torso garment 20 may further comprise neckband 34 that extends to edge 36. Torso garment 20 may comprise securing means to secure neck gaiter 70. Such securing means may be positioned anywhere thereon, such as at rear side 24, or below or adjacent to neckband 34. In the illustrated embodiment, neckband 34 comprises the securing means to secure neck gaiter 70. Such securing means includes, but is not limited to fasteners, spring snaps assemblies, hook and loop fasteners, and zipper assemblies. For illustrative purposes, the securing means in this embodiment comprises female spring snap assemblies 38 that are positioned at neckband 34. Such spring snaps assemblies may comprise as an example caps, sockets, studs, and posts (eyelets). It is noted that the securing means may be positioned on either an interior or exterior side of neckband 34. To further protect from UV radiation, present invention comprises neck gaiter 70. Neck gaiter 70 comprises exterior side 72 and interior side 80 that extend between top end 74 and bottom end 76. Neck gaiter 70 further comprises male spring snap assemblies 78 that are positioned at interior side 80. It is noted that the securing means at neck gaiter 70 may also be positioned at the exterior side 72. Furthermore, the male and female spring snap assemblies may be switched, the intension being that they connect/mate to function as securing means. As seen in FIG. 3, present invention 10 is worn by a second user with shorter arms than that of the user illustrated in FIG. 2, whereby elastic band 42 snugly and comfortably fits at, or approximately at, a user's wrist section. In a preferred embodiment, torso garment 20 and hand covers 40 are made of comfortable stretchable fabrics/materials so that thumb loop 54, and first and second finger loops 58 and 60 remain snug while worn by the user upon the thumb, index, and little finger respectively. In addition, neck gaiter 70 is also made of are made of comfortable stretchable fabrics/materials to snugly fit over the user's neck and onto a neck area of the user. Neck gaiter 70 is also sufficiently large and elastic to comfortably cover the ears, nose, and mouth if desired by the user. Such comfortable stretchable fabrics/materials include, but are not limited to spandex, cotton, cotton blends, nylon, polyesters, combinations thereof, or any other comfortable stretchable fabrics/materials having similar characteristics to said spandex, cotton, cotton blends, nylon, polyesters materials. The foregoing description conveys the best understanding of the objectives and advantages of the present invention. Different embodiments may be made of the inventive concept of this invention. It is to be understood that all matter disclosed herein is to be interpreted merely as illustrative, and not in a limiting sense. |
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claims | 1. A natural circulation boiling water reactor in operation having a chimney comprising:a cylindrical chimney shell disposed above a core in a reactor pressure vessel;a plurality of square tubes, each square tube having a lower end and an upper end, disposed so as to extend in a vertical direction in said chimney shell; anda first grid support plate, disposed in said reactor pressure vessel above said core, arranged to have a plurality of first grid holes and a plurality of second grid holes so that said first grid holes and said second grid holes are arranged alternately in rows and columns;wherein said lower ends of said square tubes are disposed and detachably supported by said second grid holes of said first grid support plate such that said second grid hole faces an opening at said lower end of said square tube, and said square tubes are not disposed at positions of said first grid holes such that said first grid holes are vacant without any of said square tubes being disposed thereat;wherein said square tubes are disposed with respect to said second grid holes in the rows and columns so as to be spaced from each other at an interval which exceeds a width of one of said square tubes;wherein a gap is formed between edges of said square tubes of said adjacent rows which face each other at least along a diagonal of a cross section of the square tubes; andwherein a two-phase flow path communicated with said core includes said first grid holes of said first grid support plate where said square tubes are not disposed and outer side surfaces of said square tubes extending upwardly from said second grid holes of said first grid support plate where said square tubes are disposed. 2. The natural circulation boiling water reactor according to claim 1,wherein each of said square tubes is a bent plate member having a welded seam which extends in a longitudinal direction of said square tube. 3. The natural circulation boiling water reactor according to claim 2, wherein a welding line for said welded seam is formed at one of the outer side surfaces of said square tube, and is positioned between ⅛ and ⅜ of the width of the one of the outer side surfaces. 4. The natural circulation boiling water reactor according to claim 1, wherein said two-phase flow path communicated with said core further includes an inside of said square tubes. 5. The natural circulation boiling water reactor according to claim 1, further comprising:a second grid support plate in said reactor pressure vessel, disposed above said first grid support plate, having a plurality of third grid holes and a plurality of fourth grid holes so that said third grid holes and said fourth grid holes are arranged alternately in rows and columns corresponding to the alternate arrangement of said first grid holes and said second grid holes of said first grid support plate, respectively;wherein said upper ends of said square tubes are detachably supported by said fourth grid holes such that said fourth grid hole faces an opening at said upper end of said square tubes. 6. The natural circulation boiling water reactor according to claim 1, further comprising a plurality of fuel assemblies disposed in the core, wherein said fuel assembles are configured to be inserted through said first and second grid holes. 7. The natural circulation boiling water reactor according to claim 1, wherein said square tubes are spaced from each other at an interval which exceeds a width of one of said square tubes in each row such that an arrangement of said square tubes disposed in one row are offset with respect to an arrangement of said square tubes disposed in an adjacent row. 8. The natural circulation boiling water reactor according to claim 1, wherein said first grid holes and said second grid holes are substantially of a same size enabling said lower ends of said square tubes to be disposed and detachably supported thereby. |
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description | This application claims the benefit of U.S. Provisional Applications Ser. Nos. 60/895,126 filed Mar. 15, 2007 and 60/921,733 filed Apr. 3, 2007. 1. Field of Use for the Invention This invention relates to the field of charged particle optics, and in particular to systems for generation of high current density shaped electron beams. 2. Description of the Related Art The use of electron beams to lithographically pattern semiconductor masks, reticles and wafers is an established technique. The different lithography strategies may be characterized by the following key parameters: beam positioning strategy; and beam shape control. There are two main approaches to the positioning of electron beams for the exposure of resist during the lithographic process: (a) Raster Scanning, where the beam is moved on a regular two-dimensional lattice pattern. This method has the advantage that the scan electronics is typically simpler, but the disadvantage is that the beam may spend large amounts of time moving across areas not needing to be exposed. In addition, in order to accomplish very precise pattern edge placement, sophisticated gray-scale and/or multiple-pass scanning may be required. (b) Vector Scanning where the beam is moved two-dimensionally directly to areas to be written. This method has the advantage of reduced time over areas not needing to be exposed, but the disadvantage of more complicated and expensive deflection electronics. Precise pattern edge placement is also easier, utilizing the beam placement capability on a 2D address grid much smaller than the beam size.Each approach is advantageous in certain circumstances, the optimum choice depending on the critical dimensions of the pattern, pattern density (% of area to be written), and also on the profile of the beam current distribution. There are two well-known approaches to the shaping of the electron beam used to expose the resist on the substrate: (a) Gaussian beams are characterized by the highest current densities (typically >2000 A/cm2) since in these systems, an image of the electron source is focused onto the substrate surface, thereby taking full advantage of the high brightness of the source. A key disadvantage of Gaussian beams is their long tails of current, stretching far outside the central beam diameter—only 50% of the beam current at the substrate falls within the FWHM of a two-dimensional Gaussian distribution. (b) Shaped Beams are formed by electron optical columns typically having several intermediate shaping apertures, combined with additional deflectors and lenses to form a focused image of the aperture(s) on the substrate surface. These systems typically have beam current densities orders-of-magnitude lower (e.g. 20-50 A/cm2) than for the Gaussian beams. An advantage of these systems is the reduced current tails outside the desired beam shape, making patterning less susceptible to process fluctuations. Another advantage is that effectively a large number of pixels may be written simultaneously since the area of the variable shaped beam may be large in comparison to a single pixel—this has the effect of increasing the writing throughput since fewer “flashes” of the electron beam are required to write a pattern. There is a need in the semiconductor industry to achieve the highest patterning throughputs, both for mask and reticle writing as well as potentially for the direct writing of wafers. Either of the two approaches to beam positioning can be combined with either of the two approaches to beam shaping, but none of these four combinations is capable of fully meeting the semiconductor industry's needs. Clearly there is a need for an electron lithography system having high throughput (at least several wafers/hour or less than an hour to write a reticle), combined with the ability to pattern very small CDs with edge placement accuracies <CD/8, as well as the simplest possible electron optical design to ensure adequate system reliability, long mean-time-between-failures (MTBF) and short mean-time-to-repair (MTTR). A third possible contribution to increasing throughput is to use multiple beams in parallel to lithographically pattern a single wafer. The challenges associated with using multiple beams include: scaling electron beam columns to fit multiple columns over a single wafer; stitching together the areas patterned by different columns; and the complexity and hardware costs associated with multiple columns. In order to achieve high throughput, there is clearly a need to have a writing system with two or three of the following characteristics: 1) multiple beams writing in parallel on the same substrate; 2) a high beam current density in a shaped beam; 3) an efficient writing strategy such as vector scanning. There is a need for a lithography system which makes best use of the above three characteristics. The present invention provides an optical column for charged-particle direct-writing which generates a high current density charged particle beam, coupled with the ability to dynamically shape the beam into non-circular profiles at the substrate being written on. According to aspects of the invention, a first embodiment of the charged particle shaped beam column includes: a charged particle source; a gun lens configured to provide a charged particle beam approximately parallel to the optic axis of the column; an objective lens configured to form the charged particle shaped beam on the surface of a substrate, wherein the disk of least confusion of the objective lens does not coincide with the surface of the substrate; an optical element with 8N poles disposed radially symmetrically about the optic axis of the column, the optical element being positioned between the condenser lens and the objective lens, wherein N is an integer greater than or equal to 1; and a power supply configured to apply excitations to the 8N poles of the optical element to provide an octupole electromagnetic field. The octupole electromagnetic field induces azimuthally-varying third-order deflections to the beam trajectories passing through the 8N-pole optical element. These beam deflections, when combined with spherical aberration in the optical system and defocus in the objective lens, induce an azimuthally-varying effective spherical aberration which causes the beam profile to deviate from circularity. By controlling the excitation of the 8N poles, it is possible to generate a square beam at the substrate, or a partially-square beam with rounded corners. The 8N-pole element can be a magnetic 8N-pole element, where the excitation is a current, or an electrostatic 8N-pole element, where the excitation is a voltage. The charged particle beam may be an electron or ion beam. The 8N pole optical element allows for a fully rotatable octupole field for N>2. The larger the value of N, the more control there is over the quadrupole and octupole fields generated. However, large values of N result in greater complexity and cost. The invention is not limited to generating square beams at the surface of the substrate. Other shapes, such as rectangles may also be generated using the structure and method of the present invention. For example, with the addition of non-octupole excitations, rectangular or parallelogram-shaped beams are possible. Further aspects of the first embodiment of the invention include a method of forming a charged particle shaped beam in a charged particle optical column. The method includes the steps of: forming a charged particle beam approximately parallel to the optic axis of the charged particle column; creating an octupole electromagnetic field to induce azimuthally dependent deflection of the charged particle beam, wherein the azimuthal angle is about the optic axis of the charged particle column, in a plane perpendicular to the optic axis; and forming a charged particle shaped beam on a substrate. A second embodiment of the present invention enables more complete control of the beam profile at the substrate. According to aspects of the invention, a second embodiment of a charged particle shaped beam column includes: a charged particle source; a gun lens configured to provide a charged particle beam approximately parallel to the optic axis of the column; an objective lens configured to form the charged particle shaped beam on the surface of a substrate; and four non-circular symmetry optical elements, each comprising 8N poles, where N is greater than or equal to 1, and N may be different for each optical element. The first 8N-pole element is excited to generate a quadrupole electromagnetic field which induces a defocusing action on the beam in a first plane (see FIG. 12A), and a focusing action on the beam in a second plane perpendicular to the first plane (see FIG. 12B). Due to this defocusing/focusing action of the first 8N-pole element, the beam profile is a first line at the second 8N-pole element. The second 8N-pole element is excited to generate a combined quadrupole and octupole electromagnetic field which induces a focusing action on the beam in the first plane and no focusing action on the beam in the second plane. Combined with the focusing action at the second 8N-pole element, the octupole excitation applied to the second 8N-pole element induces a third-order beam deflection along a first axis (the first axis being contained within the first plane). Due to this focusing action of the second 8N-pole element, the beam profile is a second line at the third 8N-pole element, where the second line is oriented 90° azimuthally with respect to the first line at the second 8N-pole element. The third 8N-pole element is excited to generate a combined quadrupole and octupole electromagnetic field which induces no focusing action on the beam in the first plane and a focusing action on the beam in the second plane. Combined with the focusing action at the third 8N-pole element, the octupole excitation induces a third-order beam deflection along a second axis (the second axis being contained within the second plane). Due to this focusing action of the third 8N-pole element, the beam profile is circular at the fourth 8N-pole element. The fourth 8N-pole element is excited to generate a combined quadrupole and octupole electromagnetic field which induces a focusing action on the beam in the first plane and a defocusing action on the beam in the second plane. Combined with the focusing action at the fourth 8N-pole element, the octupole excitation induces an azimuthally-varying third-order beam deflection. The combination of the third-order beam deflections at the second, third and fourth 8N-pole elements combines with the spherical aberration (which is azimuthally-symmetric) and defocus (also azimuthally-symmetric) to generate an azimuthally-varying beam deflection at the surface of the substrate to be written on. With proper control of the octupole excitations on the second, third and fourth 8N-poles, it is possible to generate either a square beam or a square beam with rounded corners at the surface of the substrate. The advantage of the second embodiment over the first embodiment is the more complete control of the beam profile, including the beam shape and edge acuity (i.e., the rate of current drop at the edge of the beam, measured in A/cm2 per nm of distance perpendicular to the beam edge.). The advantage of the first embodiment over the second embodiment is a simpler optical system, requiring the addition of only a single 8N-pole element. Further aspects of the present invention include a high throughput charged particle direct write lithography system including the charged particle shaped beam columns described herein. The system includes: a charged particle optical assembly configured to (1) produce a multiplicity, N, of high current density charged particle non-circular shaped-beams focused on the surface of a substrate and (2) vector scan the charged particle shaped-beams across the surface of the substrate; wherein each of the multiplicity of high current density charged particle shaped-beams has a current density, Ia, and an area A which satisfy the equations:Ia≧1000 Ampères per square centimeter;300≧N≧10;A=p2; and120>p>10 nanometers; andwherein said charged particle optical assembly includes N charged particle columns, each of the charged particle columns forming a charged particle beam, each of the charged particle columns including at least one optical element with 8N poles disposed radially symmetrically about the optic axis of the column, N being an integer greater than or equal to 1, each of the optical elements being configured to produce azimuthally dependent deflection of the corresponding charged particle beam, the azimuthal angle being about the optic axis of the corresponding charged particle column, in a plane perpendicular to the optic axis. In further aspects of the invention the parameter space for the high throughput charged particle direct write lithography system may be varied. For example, Ia≧5000 Amperes per square centimeter; 100≧N ≧10; and 120>p>20 nanometers, where A=p2. The invention disclosed herein is a charged particle beam column comprising one or more quadrupole/octupole elements which deflect the charged particle beam going down the column. The beam deflections due to the quadrupole/octupole element(s) effectively create azimuthally-varying radial deflections to the beam trajectories which, when combined with spherical aberration and defocus in the objective lens, result in forming a high current-density shaped (i.e., non-circular) beam at the substrate surface. The charged particle beam column of the invention can be either an electron beam or an ion beam column. The quadrupole/octupole optical elements can be electrostatic or magnetic elements. Many of the examples of the invention provided herein are examples of electron beam columns, with electrostatic quadrupole/octupole optical elements. However, the invention is equally applicable to ion beam columns and columns with magnetic quadrupole/octupole optical elements. Two embodiments of the present invention are described in detail herein: 1) Embodiment #1 which comprises a single additional quadrupole/octupole element (implemented using an 8N-pole optical element with combined quadrupole and octupole excitations), and 2) Embodiment #2 which comprises a quadrupole element followed by three quadrupole/octupole elements (wherein all four elements may be implemented using 8N-pole optical elements with combined quadrupole and octupole excitations).The first embodiment is described in FIGS. 6-11 and the second embodiment in FIGS. 12A-21. The relative advantages and disadvantages of the two embodiments are discussed in detail. Before describing the present invention, it is useful to first characterize the operation of a simple two-lens optical column in the absence of the present invention, as shown in FIGS. 1A-5. The present invention may be implemented in a one-lens column, but general industry practice (familiar to those skilled in the art) is to use at least two lenses in a charged particle optical column: 1) a gun (or “condenser”) lens in the electron gun to collect electrons emitted from the source (typically emitted into an expanding cone-shaped distribution) and focus their trajectories into a roughly parallel beam, which may converge to a crossover before the beam reaches the objective lens, and 2) an objective lens which focuses the electron beam generated by the gun onto a target surface. Such a two-lens column is shown in FIGS. 1A-B, for the case where there is no intermediate beam crossover between the gun lens and objective lens. The first embodiment of the present invention is applicable to columns having no intermediate crossover, as well as to columns having a single intermediate crossover. FIGS. 3-4 characterize the optical performance of the two-lens column shown in FIGS. 1A-B. The particular settings of the gun and objective lenses which generate the trajectories in FIGS. 3 and 5, and the graph in FIG. 4, have been selected for their applicability to the first embodiment of the present invention. FIG. 1A shows a schematic side view of a column employing two lenses. Electrons 121 are emitted from electron source 125 in object plane 101, which can be a thermionic source, a LaB6 emitter, a cold field emitter, a Schottky emitter, or other type of electron source as is familiar to those skilled in the art. Gun lens 102 (with focal length 111) focuses electrons 121 into an approximately parallel electron beam 122 (with radius 114) which passes down the column a distance 112 before reaching the objective lens 103 (with radius 115). Objective lens 103 (with focal length 113) focuses electrons 122 into a converging beam 123 which intersects with the surface of substrate 104 at point 126. Both lenses 102 and 103 are centered on the optical axis 127. In FIG. 1A, substrate 104 is at the paraxial focal plane of objective lens 103. FIG. 1B shows an isometric view of the two-lens column in FIG. 1A. The arrow 120 shows the direction of electron trajectories down the two-lens column. FIG. 2 shows a side view of electron trajectories converging to a focused spot on a substrate surface in a two-lens column. At the left of the graph (position 0.0 along horizontal axis 131), the beam diameter is 300000 nm (i.e., 150 μm radius) on vertical axis 132. At the resolution of this graph, electron trajectories 133 are seen to converge towards region 134 which is shown in greater detail in FIG. 3. The focal length of lens 103 (see FIGS. 1A-B) is 10.0060 mm. FIG. 3 shows a close-up side view of electron trajectories converging to a focused spot on a substrate surface in a two-lens column at region 134 in FIG. 2. The substrate is shown as a dashed line 143 at 9.9997 mm from lens 103 (see FIGS. 1A-B), having a focal length of 10.0060 mm, thus the substrate 143 is (10.0060−9.9997) mm=6.3 μm above the paraxial focal plane. The left vertical axis 142 is at 9.9970 mm from lens 103, which is (9.9997−9.9970) mm=2.7 μm above the substrate surface 143, and the top horizontal axis 141 shows the position along the optical axis 127 from lens 103. The central (on-axis) ray 144 passes from source 125, through lenses 102 and 103, and strikes the substrate 104 all on optical axis 127. Ray 145 is the farthest off-axis ray at substrate 104, but does not correspond to the farthest off-axis ray at lens 103 due to the combined effects of defocus and spherical aberration. The outer ray at lens 103 strikes substrate 104 at a radius of 146, again due to the combined effects of defocus and spherical aberration. FIG. 4 clarifies the effects of defocus combined with spherical aberration in a two-lens column. At the paraxial focal plane (10.0060 mm from lens 103), the beam displacement off-axis is due solely to spherical aberration:δX=−Csx(x2+y2)=x-axis beam displacement at paraxial focal planeδY=−Csy(x2+y2)=y-axis beam displacement at paraxial focal planewhere x and y are the beam coordinates at lens 103. Note that for electron lenses, Cs is always positive in the above formula, so δX and δY are always negative, thus spherical aberration causes the electron trajectories to cross optical axis 127 before reaching the paraxial image plane. Now if we move the substrate 104 above the paraxial image plane, we must add defocus terms to the equations for δX and δY:δX=(Δf/f)x−Csx(x2+y2)δY=(Δf/f)y−Csy(x2+y2)where f=the focal length 113 of lens 103, and Δf=the amount of defocus (i.e., the distance above the paraxial focal plane where the substrate is positioned). Clearly, for small x and y, the linear terms dominate δX and δY, but as x and/or y is increased (corresponding to rays which are not paraxial at lens 103), eventually the cubic spherical aberration terms come to dominate δX and/or δY. FIG. 4 shows a graph 153 of the radii 152 of the electron trajectories at the substrate 104 surface against the radii 151 of the trajectories at objective lens 103 in a two-lens column. The central ray strikes substrate 104 a position 154 (0 nm off-axis)—this corresponds to point 144 in FIG. 3. As the radius 151 at objective lens 103 is increased, defocus initially makes the radii at the substrate 104 increase from 0 mm at point 154 to point 155—this is region 156. For radii at objective lens 103 larger than point 155 (in region 159), spherical aberration starts to dominate and the radii 152 at the substrate 104 start to decrease, crossing the 0 nm axis at point 157 and ending up at point 158 which is on the opposite side of axis 127 (see FIGS. 1A-B) from point 155. This is a common phenomenon familiar to those skilled in the art. The curve 153 is the same for any azimuthal (i.e., angle around the axis 127) initial position of the trajectory at lens 103 since the beam is circular (see FIG. 5). Note that axis 152 includes both positive and negative numbers for the radius at the substrate 104—in this case, a negative radius corresponds to a positive radius of the same magnitude, but rotated azimuthally by 180° around the optical axis 127. FIG. 5 shows a graph of the trajectories 163 along the X-axis 161 and Y-axis 162 at the substrate 104. Since the beam-defining aperture (not shown) is round, the beam at the substrate 104 is also round. The distribution of current within the round beam is determined by the interaction of defocus and spherical aberration as illustrated in FIG. 4. Generally there is a concentration of current around the origin of the X-Y coordinate system at the substrate 104 as shown by the dark area at the center of FIG. 5. First Embodiment FIG. 6 shows a schematic side view of a first embodiment of the present invention. Electrons 218 are emitted from electron source 215 in object plane 201, which can be a thermionic source, a LaB6 emitter, a cold field emitter, a Schottky emitter, or other type of electron source as is familiar to those skilled in the art. The particular type of electron source is not part of the present invention. Gun lens 202 (with focal length 211) focuses electrons 218 into an approximately parallel electron beam 219 which passes down the column a distance 212 before reaching octupole 203. Octupole 203 may be implemented in the column using an element with 8N poles, where N=1 (an octupole), 2 (a 16-pole), . . . as is familiar to those skilled in the art. FIG. 7 shows a view of a 16-pole element (N=2). The excitation of octupole 203 is discussed in FIG. 7. Trajectories leaving octupole 203 pass a distance 213 down the column, reaching objective lens 204 (with focal length 214) which focuses electrons 220 into a converging beam 221 which intersects the surface of substrate 205 at location 216. Both lenses 202 and 204 are centered on the optical axis 217. Implementations of Octupole Elements There are a number of ways to physically implement an octupole element in an electron column. Two of these methods are illustrated in FIGS. 7 and 8. A pure octupole element (i.e., an element not also having dipole, quadrupole, hexapole, or other non-octupole excitations) is characterized by an electrostatic potential, V(x,y), with four-fold symmetry:V(x,y)=A(x4−6x2y2+y4)+B4(x3y−xy3)where A and B are constants, and x and y are the beam coordinates at the octupole element. Since the deflection of the electron trajectories passing through the octupole is proportional to the electric field, E(x,y)=−∇V(x,y), the beam deflections at the substrate, δX and δY, are:δX=K∂V(x,y)/∂x=KA(4x3−12xy2)+KB(12x2y−4y3)δY=KV(x,y)/∂y=KA(−12x2y+4y3)+KB(4x3−12xy2)Where K is a constant that depends on the beam energy passing through the octupole, the length and bore of the octupole poles, and the focal length of the objective lens. The constant A corresponds to an octupole oriented along the X- and Y-axes, while the constant B corresponds to an octupole oriented 22.5° relative to the X- and Y-axes. In the following discussion, B=0 for simplicity. For complete generality (i.e., arbitrary orientations of the shaped beam), both A and B would be non-zero. FIG. 7 shows a schematic view of an electrostatic 16-pole optical element that can be used for octupole 203 (see FIG. 6) in a first embodiment of the present invention, and for elements 1203-1206 (see FIGS. 12A-13B) in a second embodiment of the present invention. The sixteen poles 233-248 are oriented relative to the X-axis 231 and Y-axis 232 as shown. Table I shows octupole excitation voltage polarities for poles 233-248 for this orientation—note that the voltage magnitudes are all the same, only the polarities differ between poles 233-248. For the first embodiment, octupole 203 has no non-octupole excitations, thus the voltages on poles 233-248 will reflect the octupole voltages in Table I only. For the second embodiment, the octupole excitations are combined with quadrupole excitations, thus the voltages on poles 233-248 will be combinations of the octupole voltages shown in Table I with quadrupole voltages in Table V. FIG. 8 shows a schematic view of an electrostatic 8-pole (octupole) optical element that can be used as an alternative to the 16-pole element described in FIG. 7. The eight poles 253-260 are oriented relative to the X-axis 251 and Y-axis 252 as shown. Table II shows octupole excitation voltage polarities for poles 253-260 for this orientation—note that the voltage magnitudes are all the same, only the polarities differ between poles 253-260. For the first embodiment, octupole 203 has no non-octupole excitations, thus the voltages on poles 253-260 will reflect the octupole voltages in Table II only. For the second embodiment, the octupole excitations are combined with quadrupole excitations, thus the voltages on poles 253-260 will be combinations of the octupole voltages shown in Table II with quadrupole voltages in Table VI. Table III shows a comparison of the relative advantages and disadvantages of the two octupole implementations shown in FIGS. 7 and 8. The key determinant between the two implementations would be whether all orientations of the beam shape are required for patterning the substrate. In general, usually only orientations along 0° and 45° are needed, so the simpler 8-pole implementation in FIG. 8 would be preferred. If, however, all orientations are required, then it is necessary to use the more complex 16-pole implementation in FIG. 7. FIG. 9 shows the beam profile and force vectors induced by quadrupole 203 in a first embodiment of the present invention, corresponding to the case where KA<0 and B=0 in the formulas for δX and δY above. The beam profile is shown as a group of concentric circles 303-308 centered on the optical axis (X=Y=0). The X-axis 301 and the Y-axis 302 are shown in units of mm, with a maximum beam radius of 150 μm. The polarities of arrows 321-328 correspond to the 0° columns in Tables I and II. For a 45° orientation of the shaped beam, the directions of arrows 321-328 would be reversed as shown in the 45° columns in Tables I and II. TABLE IOctupole excitation strengths and polarities for generating square beamsin four orientations relative to the X- and Y-axes: 0°, 22.5°, 45°, and67.5°, with the 16-pole octupole implementation in FIG. 7 or FIG. 27. Atthese four angles, the excitation strengths on the sixteen poles 233-248 arethe same. Orientations at other angles between 0° and 90° are possibleif the excitation strengths are not the same, as is familiar to those skilled inthe art. Angles ≧90° are equivalent to angles between 0° and90° since the excitation has four-fold symmetry.Orientation of Square BeamPole #0 deg22.5 deg45 deg67.5 deg233−−++234+−−+235++−−236−++−237−−++238+−−+239++−−240−++−241−−++242+−−+243++−−244−++−245−−++246+−−+247++−−248−++− TABLE IIOctupole excitation strengths and polarities for generating asquare beam in two orientations relative to the X- and Y-axes: 0° and 45°with the 8-pole implementation in FIG. 8 or FIG. 28. At these two angles,the excitation strengths on the eight poles 253-260 are the same.Orientations at other angles are not possible with an 8-pole configuration.Angles ≧90° are equivalent to angles between 0° and90° since the excitation has four-fold symmetry.Orientationof Square BeamPole #0 deg45 deg253−+254+−255−+256+−257−+258+−259−+260+− TABLE IIIA comparison of the relative advantages and disadvantages ofthe two alternative implementations of an octupole element as shown inFIGS. 7 or 27 compared with FIGS. 8 or 28.OctupoleImplementationAdvantagesDiasadvantages16-polecapability for beam shapeincreased complexity, more(FIG. 7 & 27)orientation at any anglewires, more electronics8-polesimpler - fewer wires andbeam can only be oriented(FIG. 8 & 28)less electronicsat 0 and 45 deg FIGS. 7 or 27 compared with FIGS. 8 or 28. FIG. 10 shows a graph of the radii 342 of the electron trajectories at the substrate surface 205 against the radii 341 of the trajectories at the objective lens 204 in a first embodiment of the present invention. Comparison of FIG. 10 with FIG. 4 for a two-lens circular beam column shows that now there are two curves, 343 and 344, instead of only one (e.g., curve 153 in FIG. 4)—this is because the beam is shaped into a square by azimuthal control of the total spherical aberration, as described below. In this example, the goal is to generate a square-shaped beam with 61 nm sides, where the sides of the square-shaped beam are aligned with the X-axis 361 or Y-axis 362 in FIG. 11. Thus the distance from the center of the beam to the side is 30.5 nm (short-dashed line 350) and the distance to the corners is √2 (30.5 nm)≈43.1 nm (long-dashed line 351). By use of octupole element 203, the total deflection of the beam now has the combined effects of three terms: a) defocus, b) spherical aberration in lens 204 (equivalent to lens 103 in FIGS. 1A-B), and c) the deflection due to octupole 203 (assuming B=0):δX=(Δf/f)x−Csx(x2+y2)+KA(4x3−12xy2)δY=(Δf/f)y−Csy(x2+y2)+KA(−12x2y+4y3)The terms in these equations can be rearranged:δX=(Δf/f)x+(4KA−Cs)x3−(12KA+Cs)xy2 δY=(Δf/f)y−(12KA+Cs)x2y+(4KA−Cs)y3 If K A <0, since Cs>0, then the on-axis terms (i.e., terms with x3 and y3) are increased, while the off-axis terms (i.e., terms with x y2 and x2 y) are decreased. Curve 352 is equivalent to curve 153 in FIG. 4, corresponding to azimuthally-uniform spherical aberration. Note that the end point 353 of curve 352 is midway between the endpoint 349 of curve 344 and the endpoint 347 of curve 343—this shows the effects of adding the octupole beam deflection due to element 203. By proper choice of defocus Δf, combined with the value of A, it is possible to bring the tangent point 348 of curve 344 to match the required radius 351 of the shaped beam corner at (43.1 nm). At the same time, the tangent point 346 of curve 343 is matching the required radius 350 of the shaped beam sides (at 30.5 nm). All three curves 343, 344, and 352 start at point 345 on axis 342 (i.e., at 0 nm radius at the substrate and at 0.00 mm radius at the objective lens 204). FIG. 11 shows a graph of the trajectories 363 along the X-axis 361 and Y-axis 362 at the substrate 205 (see FIG. 6) with the use of the first embodiment of the present invention to shape the beam into a square, instead of the round beam 163 shown in FIG. 5, which would result if the octupole element 203 were inactivated (i.e., if all the poles 233-248 in FIG. 7, or all of the poles 253-260 in FIG. 8 were set to the same voltage). The sharpness of corners 365 can be controlled by adjusting the strength of octupole element 203 (i.e., adjusting the value of constant A). Curves 343 and 344 in FIG. 10 show that there is substantial overlap of the trajectories 363 in FIG. 11—this overlap can be seen from the fact that both curves 343 and 344 show two different radii at the objective lens 204 (axis 342) for the same radius at substrate 205 (axis 341) in many cases. This overlap corresponds to a “folding over” of the beam on itself, thus making the beam smaller for a given number of trajectories reaching the substrate 205. Since the number of trajectories is proportional to the total beam current, this means that the beam current density at the substrate 205 is increased compared with the case of first-order imaging (the conventional method of beam-shaping) in which there is no folding over of the trajectories at the substrate. If A is set=0, and B≠0, then a square rotated 45° to that shown in FIG. 11 would result. Note that because endpoint 347 has a larger magnitude of radius than line 350, a small number of trajectories 364 strike the substrate 205 outside the desired 61 nm square shape. This has only a minor effect since only a very small number of trajectories are in this group. The second embodiment of the present invention reduces or eliminates this effect. Second Embodiment FIGS. 12A-21 illustrate a second embodiment of the present invention. Table IV shows a comparison of the relative advantages and disadvantages of the first and second embodiments of the present invention. The second embodiment utilizes four elements 1203-1206, as shown in FIGS. 12A-B between the gun lens 1202 and the objective lens 1207. Elements 1203-1206 may be implemented using either 16-poles as in FIG. 7, or 8-poles as in FIG. 8. FIG. 12A shows a schematic side view of a second embodiment of the present invention in a plane containing two lines: a) a line midway between the +X-axis and +Y-axis, and b) the Z-axis=the optical axis−hereinafter this plane will be referred to as the (+X+Y)−Z plane. FIG. 12B shows a schematic side view of a second embodiment of the present invention in a plane containing two lines: a) a line between the −X-axis and +Y-axis, and b) the Z-axis=the optical axis−hereinafter this plane will be referred to as the (−X+Y)−Z plane. Note that this plane is perpendicular to the (+X+Y)−Z plane of FIG. 12A. FIG. 13A shows a schematic isometric view of a second embodiment of the present invention, viewed in a direction approximately perpendicular to the +X+Y plane in FIG. 12A. FIG. 13B shows a schematic isometric view of a second embodiment of the present invention in a direction 90° away from the viewing direction for FIG. 13A (the viewing direction is approximately perpendicular to the −X+Y plane in FIG. 12B). The following discussion refers to all of FIGS. 12A-13B. Electrons 1221 and 1231 are emitted from electron source 1241 in object plane 1201, which can be a thermionic source, a LaB6 emitter, a cold field emitter, a Schottky emitter, or other type of electron source as is familiar to those skilled in the art. The particular type of electron source is not part of the present invention. Gun lens 1202 (with focal length 1211) focuses electrons 1221 and 1231 into approximately parallel electron beams 1222 and 1232, respectively, of diameter 1312 which pass down the column a distance 1212 to reach quadrupole #1 1203 at a diameter 1313. In the (+X+Y)−Z plane (FIG. 12A), quadrupole #1 1203 is a diverging lens, while in the (−X+Y)−Z plane (FIG. 12B), quadrupole #1 1203 is a converging lens. The focal length of quadrupole #1 1203 is set equal to the distance 1213 between optical elements 1203 and 1204. Thus, in the (−X+Y)−Z plane (FIG. 12B), beam 1233 is brought to a focus at the center of quadrupole/octupole #2 1204. In the (+X+Y)−Z plane (FIG. 12A), beam 1223 is twice as far off-axis at quadrupole/octupole #2 1204 as at quadrupole #1 1203. The effect of quadrupole #1 1203 on beams 1222 and 1232 is shown in FIG. 14. Due to the focusing effects of quadrupole #1 1203, the beam profile at quadrupole/octupole #2 1204 is a line 1314 (seen most clearly in FIG. 13A) that is twice as long as beam diameter 1313. The effect of quadrupole/octupole #2 1204 on beams 1223 and 1233 is shown in FIG. 15. Because in the (−X+Y)−Z plane the beam is on-axis, there is no focusing effect due to quadrupole/octupole #2 1204. In the (+X+Y)−Z plane, the beam 1223 is strongly focused towards optical axis 1240, generating converging beam 1224. In the (−X+Y)−Z plane, the beam 1234 diverges away from optical axis 1240. In the example shown here, the relationships between the spacings of elements 1203-1206 are as follows:Spacing 1214=2 (spacing 1213)=2 (spacing 1215)Spacing 1211 is the focal length of gun lens 1202, while spacing 1217 is approximately the focal length of objective lens 1207. As long as the beam is assumed parallel after lens 1202, spacing 1212 is unimportant. As long as the beam is parallel after quadrupole/octupole #4 1206, spacing 1216 is also unimportant. Midway between quadrupole/octupole #2 1204 and quadrupole/octupole #3 1205, the beam is circular with a diameter 1318. Due to the focusing effects of quadrupole/octupole #2 1204, the beam profile at quadrupole/octupole #3 1205 is a line 1315 (seen most clearly in FIG. 13B) that is equal in length to line 1314, but rotated 90° azimuthally. The effect of quadrupole/octupole #3 1205 on beams 1224 and 1234 is shown in FIG. 16. Because in the (+X+Y)−Z plane (FIG. 12A) the beam is on-axis, there is no focusing effect due to quadrupole/octupole #3 1205 and the beam 1225 diverges away from optical axis 1240. In the (−X+Y)−Z plane (FIG. 12B), the beam 1234 is strongly focused towards optical axis 1240, generating converging beam 1235. Due to the focusing effects of quadrupole/octupole #3 1205, the beam profile at quadrupole/octupole #4 1206 is a circle 1316. The effect of quadrupole/octupole #4 1206 on beams 1225 and 1235 is shown in FIG. 17. Because in both the (+X+Y)−Z and (−X+Y)−Z planes the beam is off-axis, there is a lens effect for all positions on the beam diameter 1316. In the (+X+Y)−Z plane (FIG. 12A), beam 1225 is focused towards optical axis 1240, generating parallel beam 1226. In the (−X+Y)−Z plane (FIG. 12B), the beam 1235 is focused away from optical axis 1240, generating parallel beam 1236. The parallel beams 1226 and 1236 reach objective lens 1207 on circle 1317, where all electrons are focused towards the substrate 1208 at point 1242. In the preceding discussion, only the first-order focusing effects of elements 1203-1206 have been discussed—these are the optical effects of the quadrupole excitations of elements 1203-1206. In order to shape the beam, however, it is necessary to add octupole excitations to elements 1204-1206, as will be described in FIGS. 15-17. TABLE IVComparison of the advantages and disadvantages of the first andsecond embodiments of the present invention.EmbodimentAdvantagesDiasadvantagesFirstsimpler - no addedsome rays outside square,components, only additionless ability to fine-tuneof octupole excitation tocorner roundingexisting deflectors (8N-poles)Secondno rays outside square,more complex - need toability to fine-turn corneradd some 8N-poleroundingelements to column design,added wires andelectronics FIG. 14 shows the beam profile and force vectors induced by Quadrupole #1 1203 in a second embodiment of the present invention. The beam profile is shown as a group of concentric circles 1403-1408 centered on the optical axis (X=Y=0). A pure quadrupole element (i.e., an element not also having dipole, hexapole, octupole, or other non-quadrupole excitations) is characterized by an electrostatic potential, V(x,y):V(x,y)=C(x2−y2)+D2xy where C and D are constants, and x and y are the beam coordinates at the quadrupole element. Since the deflection of the electron trajectories passing through the quadrupole is proportional to the electric field, E(x,y)=−∇V(x,y), the beam deflections at the next element (e.g., at element 1204 due to deflection by element 1203, etc.), δXo, and δYo, are:δXo=Q∂V(x,y)/∂x=QC2x+QD2y δYo=Q∂V(x,y)∂y=−QC2y+QD2x where Q is a constant that depends on the beam energy passing through the quadrupole and the length and bore of the quadrupole poles. The constant C corresponds to a quadrupole oriented along the X- and Y-axes, while the constant D corresponds to an quadrupole oriented 45° relative to the X- and Y-axes. In the following discussion, C=0, corresponding to the requirement to generate line foci 1314 and 1315 oriented 45° relative to the X- and Y-axes. For complete generality (i.e., arbitrary orientations of the shaped beam), both C and D would be non-zero. Use of quadrupoles to shape beams down an electron beam column is familiar to those skilled in the art. The beam profile at quadrupole #1 1203 is shown as a group of concentric circles 1403-1408 centered on the optical axis (X=Y=0). The X-axis 1401 and the Y-axis 1402 are shown in units of mm, with a maximum beam radius of 150 μm. The four double arrows 1410-1413 represent forces on the beam due to the quadrupole excitation as shown in the columns for 45° in Table V (for the 16-pole in FIG. 7 or FIG. 27) and Table VI (for the 8-pole in FIG. 8 or FIG. 28). FIG. 15 shows the beam profile and force vectors induced by Quadrupole/Octupole #2 1204 in a second embodiment of the present invention. The X-axis 1501 and the Y-axis 1502 are shown in units of mm, with a maximum beam distance off-axis of 300 μm, or twice the 150 μm radius in FIG. 14, as described in FIGS. 12A-13B, above. The two double arrows 1510 and 1511 show the first-order converging effects of the quadrupole excitation of quadrupole/octupole #2 1204. The two single arrows 1521 and 1522 show the third-order diverging effects of the octupole excitation of quadrupole/octupole #2 1204. Note that both the quadrupole and octupole effects act only along the (+X+Y)-axis where the beam has a non-zero radius since in all cases the beam deflection is a function of the beam radius. Quadrupole #1 1203 generates a line beam at quadrupole/octupole #2 1204 so that the octupole excitation of quadrupole/octupole #2 1204 can act on the beam only along the (+X+Y)−direction, thereby adjusting the sharpness of two diagonal corners of the final shaped beam at the substrate 1208. TABLE VQuadrupole excitation strengths and polarities for generating linebeams in eight orientations relative to the X- and Y-axes: 0° to 157.5°in steps of 22.5°, with the 16-pole implementation in FIG. 7 or 27.Two people strength magnitudes are shown for these orientationangles: 1.000 and 0.414 = (√2 − 1). Orientations at other anglesbetween 0° and 180° are possible with other pole strengths as isfamiliar to those skilled in the art. Angles ≧180° areequivalent to angles between 0° and 180° since theexcitation has two-fold symmetry.Orientation of Line Focus0.022.545.067.590.0112.5135.0157.5Pole #degdegdegdegdegdegdegdeg233+++a−a−−−a+a234+a+++a−a−−−a235−a+a+++a−a−−236−−a+a+++a−a−237−−−a+a+++a−a238−a−−−a+a+++a239+a−a−−−a+a++240++a−a−−−a+a+241+++a−a−−−a+a242+a+++a−a−−−a243−a+a+++a−a−−244−−a+a+++a−a−245−−−a+a+++a−a246−a−−−a+a+++a247+a−a−−−a+a++248++a−a−−−a+a+a = cos(2 * 33.75 deg)/cos (2 * 11.25 deg) = 0.414 Table V. Quadrupole excitation strengths and polarities for generating line beams in eight orientations relative to the X- and Y-axes: 0° to 157.5° in steps of 22.5°, with the 16-pole implementation in FIG. 7 or 27. Two pole strength magnitudes are shown for these orientation angles: 1.000 and 0.414=(√2−1). Orientations at other angles between 0° and 180° are possible with other pole strengths as is familiar to those skilled in the art. Angles ≧180°are equivalent to angles between 0° and 180° since the excitation has two-fold symmetry. TABLE VIQuadrupole excitation strengths and polarities for generating line beamsin four orientations relative to the X- and Y-axes: 0°, 45°, 90° and135° with the 8-pole implementation in FIG. 8 or 28. At these fourangles, the excitation strengths on the eight poles 253-260 are the same.Orientations at other angles are between 0° and 180° are not possiblewith an 8-pole implementation. Angles >180° are equivalent to anglesbetween 0° and 180° since the excitation has two-fold symmetry.Orientation of Line FocusPole #0 deg45 deg90 deg135 deg253+0−02540+0−255−0+02560−0+257+0−02580+0−259−0+02600−0+ Table VI. Quadrupole excitation strengths and polarities for generating line beams in four orientations relative to the X- and Y-axes: 0°, 45°, 90° and 135° with the 8-pole implementation in FIG. 8 or 28. At these four angles, the excitation strengths on the eight poles 253-260 are the same. Orientations at other angles between 0° and 180° are not possible with an 8-pole implementation. Angels >180° are equivalent to angles between 0° and 180° since the excitation has two-fold symmetry. FIG. 16 shows the beam profile and force vectors induced by Quadrupole/Octupole #3 1205 in a second embodiment of the present invention. The X-axis 1601 and the Y-axis 1602 are shown in units of mm, with a maximum beam distance off-axis of 300 μm, or equal to the maximum beam distance off-axis in FIG. 15, as described in FIGS. 12A-13B, above. The two double arrows 1610 and 1611 show the first-order converging effects of the quadrupole excitation of quadrupole/octupole #3 1205. The two single arrows 1621 and 1622 show the third-order diverging effects of the octupole excitation of quadrupole/octupole #3 1205. Note that both the quadrupole and octupole effects act only along the (−X+Y)-axis where the beam has a non-zero radius since in all cases the beam deflection is a function of the beam radius. Intuitively, quadrupole/octupole #2 1204 generates a line beam at quadrupole/octupole #3 1205 so that the octupole excitation of quadrupole/octupole #3 1205 can act on the beam only along the (−X+Y)-direction, thereby adjusting the sharpness of two diagonal corners of the final shaped beam at the substrate 1208 (the two corners not adjusted by quadrupole/octupole #2 1204). FIG. 17 shows the beam profile and force vectors induced by Quadrupole/Octupole #4 1206 in a second embodiment of the present invention. The beam profile is shown as a group of concentric circles 1703-1708 centered on the optical axis (X=Y=0). The X-axis 1701 and the Y-axis 1702 are shown in units of mm, with a maximum beam radius of 150 μm, equal to the radius in FIG. 14, as described in FIGS. 12A-13B, above. The four double arrows 1710-1713 show the first-order converging effects of the quadrupole excitation of quadrupole/octupole #4 1206. The eight single arrows 1721-1728 show the third-order converging and diverging effects of the octupole excitation of quadrupole/octupole #4 1206. Note that both the quadrupole and octupole effects act in all directions azimuthally since the beam has non-zero radius for all trajectories. Intuitively, quadrupole/octupole #3 1205 generates a circular beam at quadrupole/octupole #4 1206 so that the octupole excitation of quadrupole/octupole #3 1205 can act on the beam in all directions, in the same way that octupole 203 acts on the beam in the first embodiment. The combined effects of elements 1204-1206 is to shape the beam into a square, at the substrate 1208, but with increased adjustability of corner sharpness compared with the first embodiment due to the additional octupole excitations in elements 1204-1205 (see comparison in Table IV). FIG. 18 shows a graph of the radii 1802 of the electron trajectories at the substrate surface 1802 against the radii 1801 of the trajectories at the objective lens in a second embodiment of the present invention, in which only quadrupole/octupole #4 1206 has an octupole excitation—this example is only for illustration, and basically corresponds to operating in a mode similar to the first embodiment. In this case, elements 1204 and 1205 function only as quadrupoles. Comparison of FIG. 18 with FIG. 4 for a circular beam column shows that now there are two curves, 1803 and 1804, instead of only one—this is because the beam is shaped into a square by azimuthal control of the spherical aberration, as described below. The goal is to generate a square-shaped beam with 66 nm sides. Thus the distance from the center of the beam to the side is 33 nm (short dashed line 1805) and the distance to the corners is √2 (33 nm)≈46.7 nm (long dashed line 1806). By use of quadrupole/octupole element 1206, the total deflection of the beam now has the combined effects of three terms, defocus, spherical aberration in lens 1207 (equivalent to lens 103 in FIGS. 1A-B), and the deflection due to quadrupole/octupole 1206, as was described above for the first embodiment. In FIG. 18, the octupole excitation of quadrupole/octupole 1206 has intentionally been set low to leave the corners of the beam in FIG. 19 rounded. In many applications where sidewall coverage during deposition is an issue, it is preferable not to generate a beam with sharp corners, since the resulting etched square hole (typically a contact or via) would be difficult or impossible to completely fill with conductive material (such as tungsten, copper, aluminum, etc.). Because of the corner rounding in FIG. 19, the minimum 1807 of curve 1803 does not quite reach the desired 33 nm beam radius of 1805 for the sides of the square beam. Similarly, the minimum 1808 of curve 1804 does not quite reach the desired 46.7 nm radius of 1805 for the corners of the square beam. Note that axis 1802 includes both positive and negative numbers for the radius at the substrate 1208—in this case, a negative radius corresponds to a positive radius of the same magnitude, but rotated azimuthally by 180° around the optical axis 1240. FIG. 19 shows a graph of the trajectories 1813 along the X-axis 1811 and Y-axis 1812 at the substrate 1208 with the use of the second embodiment of the present invention to shape the beam into a square with rounded corners 1814 in which only quadrupole/octupole #4 1206 has an octupole excitation. As described in FIG. 18, in many applications, some rounding of the corners of the beam may be advantageous to improve sidewall coverage during deposition into the contact or via. Curves 1807 and 1808 in FIG. 18 show that there is substantial overlap of the trajectories 1813 in FIG. 19—this overlap can be seen from the fact that both curves 1807 and 1808 show two different radii at the objective lens 1207 (axis 1802) for the same radius at substrate 1208 (axis 1801) in many cases. This overlap corresponds to a “folding over” of the beam on itself, thus making the beam smaller for a given number of trajectories reaching the substrate 1207—this is the same phenomenon seen in FIGS. 10-11. Since the number of trajectories is proportional to the total beam current, this means that the beam current density at the substrate 1208 is increased compared with the case of first-order imaging (the conventional method of beam-shaping) in which there is no folding over of the trajectories at the substrate. FIG. 20 shows a graph of the radii 1902 of the electron trajectories at the substrate 1208 against the radii 1901 of the trajectories at the objective lens in a second embodiment of the present invention, in which elements 1204-1206 all have octupole excitations as shown in FIGS. 15-17. Comparison of FIG. 19 with FIG. 4 for a circular beam column shows that now there are two curves, 1903 and 1904, instead of only one—this is because the beam is shaped into a square by azimuthal control of the total spherical aberration, as described below. The goal is to generate a square-shaped beam with 66 nm sides. Thus the distance from the center of the beam to the side is 33 nm (short dashed line 1905) and the distance to the corners is √2 (33 nm)≈46.7 nm (long dashed line 1906). By use of quadrupole/octupole elements 1204-1206, the total deflection of the beam now has the combined effects of three terms: defocus, spherical aberration in lens 1207 (equivalent to lens 103 in FIGS. 1A-B), and the deflection due to the octupole excitations in quadrupole/octupole elements 1204-1206. In FIG. 19, the effects of elements 1204-1207 have combined to generate the square beam in FIG. 21 which has no rays outside the desired square beam profile. The minimum 1907 of curve 1903 is tangent to the desired side radius 33 nm 1905. The minimum 1908 of curve 1904 is tangent to the desired corner radius 46.7 nm 1906. Note that axis 1902 includes both positive and negative numbers for the radius at the substrate 1208—in this case, a negative radius corresponds to a positive radius of the same magnitude, but rotated azimuthally by 180° around the optical axis 1240. FIG. 21 shows a graph of the trajectories 1913 along the X-axis 1911 and Y-axis 1912 at the substrate 1208 with the use of the second embodiment of the present invention to shape the beam into a square with corners 1914 in which elements 1204-1206 all have octupole excitations as shown in FIGS. 15-17. The corners 1914 of the beam are now sharp, which may be useful for lithography applications with substantial blurring in the resist—in these cases, the resist profile must be as sharp as possible to achieve the best final etched shape in the substrate 1208. Curves 1907 and 1908 in FIG. 20 show that there is substantial overlap of the trajectories 1913 in FIG. 21—this overlap can be seen from the fact that both curves 1907 and 1908 show two different radii at the objective lens 1207 (axis 1902) for the same radius at substrate 1208 (axis 1901) in many cases. This overlap corresponds to a “folding over” of the beam on itself, thus making the beam smaller for a given number of trajectories reaching the substrate 1207—this is the same phenomenon seen in FIGS. 10-11 and FIGS. 18-19. Since the number of trajectories is proportional to the total beam current, this means that the beam current density at the substrate 1208 is increased compared with the case of first-order imaging (the conventional method of beam-shaping) in which there is no folding over of the trajectories at the substrate. FIG. 22 shows a Venn diagram illustrating the interactions of the three contributions to the system throughput: 1) Multiple beam column assembly (circle 2201 enclosing areas 2204, 2210, 2211, and 2213)—a column assembly which can produce multiple electron beams is described in U.S. Pat. No. 6,943,351 B2, “Multiple Column Charged Particle Optics Assembly” issued Sep. 13, 2005, incorporated by reference herein. Clearly, increasing the number of beams which are simultaneously writing on a substrate will lead to a nearly-proportional increase in writing throughput. The multiple beam column technology described in the reference may be applied to the generation of both one- and two-dimensional arrays of beams, with inter-beam spacings in the range of 30 mm in X-Y, where X and Y are the coordinates in the plane of the substrate. Typical arrays of beams might comprise up to 10 beams in a line or 10×10 beams in a two-dimensional array. Area 2204 represents a system with a multiple beam column assembly using conventional low current density beam shaping and raster scanning. 2) High current density shaped beams (circle 2202 enclosing areas 2205, 2210, 2212, and 2213)—one method for achieving high current density shaped beams is the present invention. Another method for achieving high current density shaped beams is described in U.S. Patent Application Publication No. 2006/0145097 A1, “Optics for Generation of High Current Density Patterned Charged Particle Beams” filed Oct. 7, 2004, incorporated by reference herein. Both methods are capable of being implemented in the multiple beam column assembly described in the section above. The key requirement for this is the need for each column to fit within the small available X-Y footprint (typically, approximately 30 mm×30 mm) within the multiple beam column assembly. This requirement for a small column footprint generally precludes the use of complex columns with many lenses, apertures and deflectors, as are commonly used in the production of lower current density shaped beams as is familiar to those skilled in the art. The increase in throughput due to increased current density in the beam is almost proportional to the magnitude of the current density increase, assuming that blanking times between successive flashes are reasonably short compared to the flash (i.e., writing) times. In the beam shaping methods described above, current density increases of 25 to >50 times over the conventional beam shaping approaches are possible. Area 2205 represents a system with a single column using a high current density shaped beam and raster scanning. 3) Vector scanning (circle 2203 enclosing areas 2206, 2211, 2212, and 2213)—the third contribution to throughput comes from the method of deflecting the beam around on the substrate. There are two widely-used scanning methods: 1) raster-scanning where the beam always traverses an X-Y pattern and is blanked on/off to write the pattern, and 2) vector scanning where the beam is moved directly from the position of a flash to the position of the next flash. The raster approach has the benefits of greater electronic simplicity at the expense of slower writing since the beam spends a lot of time over regions not to be written (where the beam is blanked). The vector scanning approach is more complex electronically, but has the substantial benefit of reducing writing times since the beam needs to be blanked a smaller percentage of the overall writing time. Depending on the pattern density, throughput increases due to vector scanning may range from 2× to 5× compared with raster scanning. Area 2206 represents a single column system using a low current density shaped beam and vector scanning (this is the prior art shaped beam approach). Clearly to obtain the largest increases in writing throughputs, it is advantageous to combine two or all three of these contributions in one system. There are four possibilities: 1) Multiple beam column assembly with high current density shaped beams using raster scanning (area 2210)—the throughput advantage here is the product of the number of columns (10-100×) and the current density increase (25-50×)—giving an overall potential throughput increase of (250-5000×). 2) Multiple beam column assembly with low current density shaped beams and vector scanning (area 2211)—the throughput advantage here is the product of the number of columns (10-100×) and the vector scanning throughput increase (2-5×)—giving an overall potential throughput increase of (20-500×). 3) Single beam column with a high current density shaped beam and vector scanning (area 2212)—the throughput advantage here is the product of the current density increase (25-50×) and the vector scanning throughput increase (2-5×)—giving an overall potential throughput increase of (50-250×). 4) Multiple beam column assembly with a high current density shaped beam and vector scanning (area 2213)—this represents the ultimate throughput improvement situation, since the advantage here is the product of the number of columns (10-100×), the current density increase (25-50×), and the vector scanning throughput increase (2-5×)—giving an overall potential throughput increase of (500-25000×). Some examples of the parameters for combinations of multiple beam columns, high current density shaped beams and vector scanning to specify a high throughput lithography system of the invention are given below. A first example is a system with a multiplicity, N, of columns, each with a high current density charged particle shaped-beam which has a current density, Ia, and an area A, at the surface of the substrate, which satisfy the equations:Ia≧1000 Ampères per square centimeter;300≧N≧10;A=p2; and120>p>10 nanometers. A second example is a system with a multiplicity, M, of columns, each with a high current density charged particle shaped-beam which has a current density, Ib, and an area B, at the surface of the substrate, which satisfy the equations:Ib>5000 Ampères per square centimeter;100≧M≧10;B=q2; and120>q>20 nanometers. FIG. 23 shows a schematic circuit diagram of drive electronics for the 16-pole element in FIG. 7 used for element 203 in the first embodiment. Since element 203 only requires an octupole excitation, the voltages on poles 233-248 are driven by octupole driver 2302 (providing four signals: +Oct1, +Oct2, −Oct1, and −Oct2). Connections to the 16 poles 233-248 are as shown. The 4-fold symmetry inherent in an octupole excitation means that each of the four octupole signals is connected to four poles spaced 90° apart azimuthally around the optical axis. For example, signal +Oct1 connects to poles 233, 237, 241, and 245. Signals Oct1 and Oct2 are determined by the required rotation angle, θ, for the shaped beam. Table I illustrates some representative values for the voltages on poles 233-248 for four different orientations of a shaped beam. The general formulas for the voltage signals are:Oct1=A cos [4θ+45°]Oct2=A cos [4θ+135°]where A<0 is a particular voltage determined by the column optics design. Note that any rotation angle θ>90° is equivalent to an angle between 0° and 90° due to the 4 θ term. FIG. 24 shows a schematic circuit diagram of drive electronics for the 8-pole element in FIG. 8 used for element 203 in the first embodiment. Since element 203 only requires an octupole excitation, the voltages on poles 253-260 are driven by octupole driver 2402 (providing two signals: +Oct and −Oct). Connections to the 8 poles 253-260 are as shown. The 4-fold symmetry inherent in an octupole excitation means that each of the two octupole signals is connected to four poles spaced 90° apart azimuthally around the optical axis. For example, signal +Oct connects to poles 253, 255, 257, and 259. Signal Oct is determined by the required rotation angle, θ, for the octupole excitation of the 8-pole element, as is familiar to those skilled in the art. Table II illustrates some representative values for the voltages on poles 253-260 for two different orientations of a shaped beam. Since an 8-pole element can only generate two orientations of an octupole electrostatic field (θ=0° and 45°), the general formula for the voltage signal is: Oct = A cos [ 4 θ ] = A ( for θ = 0 ° ) or = - A ( for θ = 45 ° ) where A<0 is a particular voltage determined by the column optics design. Note that any rotation angle θ>90° is equivalent to an angle between 0° and 90° due to the 4θ term. FIG. 25 shows a schematic circuit diagram of drive electronics for the 16-pole element in FIG. 7 used for elements 1203-1206 in the second embodiment. Since elements 1204-1206 require both quadrupole and octupole excitations (element 1203 is a pure quadrupole), the voltages on poles 233-248 are driven by both quadrupole driver 2501 (providing eight signals: +Q1, +Q2, +Q3, +Q4, −Q1, −Q2, −Q3, and −Q4) and by octupole driver 2502 (providing four signals: +Oct1, +Oct2, −Oct1, and −Oct2). Connections to the 16 poles 233-248 are as shown. The two-fold symmetry inherent in a quadrupole excitation means that each of the eight quadrupole signals is connected to two poles spaced 180° apart azimuthally around the optical axis. For example, signal +Q1 connects to poles 233 and 241. Signals Q1, . . . , Q4 are determined by the required rotation angle, θ, for the shaped beam. Table V illustrates some representative values for the quadrupole voltages on poles 233-248 for eight different orientations of a line focus. Note that the orientation angles for the line foci are different from the orientation angle for the shaped beam. For example, a shaped beam with a rotation angle θ would require the following line focus rotation angles (see FIGS. 12A-13B and Table V): Element 1203: excitation has a θ+45° rotation—gives a line focus at θ+45° at element 1204 Element 1204: excitation has a θ+135° rotation—gives a line focus at θ+135° at element 1205 Element 1205: excitation has a θ+45° rotation—gives a round beam at element 1206 Element 1206: excitation has a θ+135° rotation—gives a parallel round beam entering lens 1207 The 4-fold symmetry inherent in an octupole excitation means that each of the four octupole signals is connected to four poles spaced 90° apart azimuthally around the optical axis. For example, signal +Oct1 connects to poles 233, 237, 241, and 245. Signals Oct1 and Oct2 are determined by the required rotation angle, θ, for the octupole excitation of the 16-pole element, as is familiar to those skilled in the art. Table I illustrates some representative values for the voltages on poles 233-248 for four different orientations of a square beam. The general formulas for the voltage signals are:Oct1=A cos[4θ+45°]Oct2=A cos [4θ+135°]where A<0 is a particular voltage determined by the column optics design. Note that any rotation angle θ>90° is equivalent to an angle between 0° and 90° due to the 4 θ term. Additive elements 2511-2518 combine the quadrupole and octupole voltages derived above. Additive elements 2511-2518 could be op-amp circuits if Q1-Q4 and Oct1-Oct2 are analog signals, or they could be digital circuitry if Q1-Q4 and Oct1-Oct2 are digital signals. In the latter case, additive elements 2511-2518 would also perform a digital-to-analog conversion to generate final (analog) drive voltages for poles 233-248. FIG. 26 shows a schematic circuit diagram of drive electronics for the 8-pole element in FIG. 8 used for elements 1203-1206 in the second embodiment. Since elements 1204-1206 require both quadrupole and octupole excitations (element 1203 is a pure quadrupole), the voltages on poles 233-248 are driven by both quadrupole driver 2601 (providing four signals: +Q1, +Q2, −Q1, and −Q2) and by octupole driver 2602 (providing two signals: +Oct and −Oct1). Connections to the eight poles 253-260 are as shown. The two-fold symmetry inherent in a quadrupole excitation means that each of the four quadrupole signals is connected to two poles spaced 180° apart azimuthally around the optical axis. For example, signal +Q1 connects to poles 253 and 257. Signals Q1 and Q2 are determined by the required rotation angle, θ, for the shaped beam. Table VI illustrates some representative values for the quadrupole voltages on poles 253-260 for four different orientations of a line focus. Note that the orientation angles for the line foci are different from the orientation angle for the shaped beam. For example, a shaped beam with a rotation angle θ would require the following line focus rotation angles (see FIGS. 12A-13B and Table VI): Element 1203: excitation has a θ+45° rotation—gives a line focus at θ+45° at element 1204 Element 1204: excitation has a θ+135° rotation—gives a line focus at θ+135° at element 1205 Element 1205: excitation has a θ+45° rotation—gives a round beam at element 1206 Element 1206: excitation has a θ+135° rotation—gives a parallel round beam entering lens 1207The 4-fold symmetry inherent in an octupole excitation means that each of the two octupole signals is connected to four poles spaced 90° apart azimuthally around the optical axis. For example, signal +Oct connects to poles 253, 255, 257, and 259. Signal Oct is determined by the required rotation angle, θ, for the octupole excitation of the 8-pole element, as is familiar to those skilled in the art. Table II illustrates some representative values for the voltages on poles 253-260 for two different orientations of a square beam. Since an 8-pole element can only generate two orientations of an octupole electrostatic field (θ=0° and 45°), the general formula for the voltage signal is: Oct = A cos [ 4 θ ] = A ( for θ = 0 ° ) or = - A ( for θ = 45 ° ) where A<0 is a particular voltage determined by the column optics design. Note that any rotation angle θ>90° is equivalent to an angle between 0° and 90° due to the 4 θ term. Additive elements 2611-2618 combine the quadrupole and octupole voltages derived above. Additive elements 2611-2618 could be op-amp circuits if Q1, Q2 and Oct are analog signals, or they could be digital circuitry if Q1, Q2 and Oct are digital signals. In the latter case, additive elements 2611-2618 would also perform a digital-to-analog conversion to generate final (analog) drive voltages for poles 253-260. FIG. 27 shows a schematic view of a magnetic 16-pole optical element that can be used for octupole 203 (see FIG. 6) in a first embodiment of the present invention, and for elements 1203-1206 (see FIGS. 12A-13B) in a second embodiment of the present invention. The sixteen magnetic poles 2733-2748 are oriented relative to the X-axis 2731 and Y-axis 2732 as shown. The operation of magnetic 16-pole optical elements is essentially equivalent to the operation of electrostatic 16-pole optical elements, with the exception that the gaps in a magnetic 16-pole element are equivalent to the poles in an electrostatic 16-pole element. This can be seen from the fact that electrons are deflected perpendicularly to a magnetic field but are deflected parallel to an electric field. Each magnetic pole in FIG. 27 is fabricated from a material with a high magnetic permeability and has a corresponding excitation coil, for example pole 2733 is excited by coil 2713 which surrounds pole 2733 next to flux return ring 2702. Identical considerations apply to poles 2734-2748 with excitation coils 2714-2728, respectively. The purpose of flux return ring 2702 is to connect together the magnetic flux generated by coils 2713-2728 to avoid excessive stray flux from adversely affecting the electron beam in parts of the column away from the 16-pole optical element. One polarity of current in an excitation coil (e.g., 2713) will make the corresponding pole (pole 2733) a North pole, while the opposite current polarity will make the corresponding pole (pole 2733) a South pole, as is familiar to those skilled in the art of magnetic deflectors. It is also possible to avoid the use magnetic materials and fabricate the 16-pole optical element using only shaped coils. This approach has the advantage of avoiding hysteresis in the magnetic poles 2733-2748 and flux return ring 2702, but with the disadvantage of requiring much higher excitation currents in coils 2713-2728. FIG. 28 shows a schematic view of a magnetic 8-pole (octupole) optical element that can be used as an alternative to the magnetic 16-pole element described in FIG. 27. The eight poles 2853-2860 are oriented relative to the X-axis 2851 and Y-axis 2852 as shown. As for FIG. 27, the operation of magnetic 8-pole optical elements is essentially equivalent to the operation of electrostatic 8-pole optical elements, with the exception that the gaps in a magnetic 8-pole element are equivalent to the Doles in an electrostatic 8-pole element. Each magnetic pole in FIG. 28 is fabricated from a material with a high magnetic permeability and has a corresponding excitation coil, for example pole 2853 is excited by coil 2813 which surrounds pole 2853 next to flux return ring 2802. Identical considerations apply to poles 2854-2860 with excitation coils 2814-2820, respectively. The purpose of flux return ring 2802 is to connect together the magnetic flux generated by coils 2813-2820 to avoid excessive stray flux from adversely affecting the electron beam in parts of the column away from the 8-pole optical element. One polarity of current in an excitation coil (e.g., coil 2813) will make the corresponding pole (pole 2853) a North pole, while the opposite current polarity will make the corresponding pole (pole 2853) a South pole, as is familiar to those skilled in the art of magnetic deflectors. It is also possible to avoid the use magnetic materials and fabricate the 16-pole optical element using only shaped excitation coils. This approach has the advantage of avoiding hysteresis in the magnetic poles 2833-2740 and flux return ring 2802, but with the disadvantage of requiring much higher excitation currents in coils 2813-2820. Table III shows a comparison of the relative advantages and disadvantages of the two octupole implementations shown in FIGS. 27 and 28. The key determinant between the two implementations would be whether all orientations of the beam shape are required for patterning the substrate. In general, usually only orientations along 0° and 45° are needed, so the simpler 8-pole implementation in FIG. 28 would be preferred. If, however, all orientations are required, then it is necessary to use the more complex 16-pole implementation in FIG. 27. The second embodiment is discussed herein with either four electrostatic 8N-pole optical elements or four magnetic 8N-pole optical elements. It is also possible to implement the second embodiment with a combination of 1-3 electrostatic 8N-pole optical elements and 1-3 magnetic 8N-pole optical elements, providing that there is a total of four 8N-pole optical elements. Both the first and second embodiments may be implemented using combined electrostatic/magnetic 8N-pole optical elements, thereby enabling partial or complete correction for chromatic aberrations in the first- and third-order deflections—the use of combined electrostatic and magnetic optical elements for chromatic aberration correction is familiar to those skilled in the art. The second embodiment may also be implemented using a configuration in which the first 8N-pole optical element has combined quadrupole/octupole excitations instead of, or in addition to, the combined quadrupole/octupole excitation on the fourth 8N-pole optical element. An advantage of this configuration is that two weaker octupole excitations (requiring may be used instead of the single, stronger, octupole excitation on the fourth 8N-pole optical element described above. A disadvantage of this configuration is that more complex electronics is required to drive the first 8N-pole optical element since it is required to generate both quadrupole and octupole fields, instead of only the quadrupole field described above. TABLE VIIParameters assumed for the calculations modeling the first embodiment.ParameterLensesOctupolesFocal length lens 202arbitraryStrength Octupole 203−0.0002501/mm{circumflex over ( )}3Focal Length Lens 20410.006mmSpherical Aberration Lens 2040.0031/mm{circumflex over ( )}3Distance Lens 204 to9.9997mmSubstrate 205 TABLE VIIIParameters assumed for the calculations modeling the second embodiment.Quadrupoles/ParameterLensesOctupolesFocal length lens 1202arbitraryStrength Quadrupole #1 12030.066671/mm0.0000001/mm{circumflex over ( )}3Strength Quadrupole/−0.066671/mm−0.0000091/mm{circumflex over ( )}3Octupole #2 1204Strength Quadrupole/0.066671/mm−0.0000091/mm{circumflex over ( )}3Octupole #3 1205Strength Quadrupole/−0.066671/mm−0.0001801/mm{circumflex over ( )}3Octupole #4 1206Focal Length Lens 120710.006mmSpherical Aberration Lens0.0031/mm{circumflex over ( )}31207Distance Lens 1207 to9.9997mmSubstrate 1208 |
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description | This application claims the benefit of Korean Patent Application No. 2006-103059, filed Oct. 23, 2006, the disclosure of which is hereby incorporated herein by reference in its entirety. 1. Field of the Invention The present invention relates to a system and method for controlling operation of semiconductor equipment. More particularly, this invention relates to a system and method of using semiconductor equipment to verify the contents of a process recipe and perform a semiconductor device manufacturing process using an appropriate recipe. 2. Description of Related Art Generally, a semiconductor device manufacturing method includes a deposition process in which a material layer is formed on a semiconductor substrate, a photolithography process in which a mask layer is formed on the material layer and the mask layer is patterned to form a mask pattern, an etching process in which the material layer is etched using the mask pattern as an etching mask, an ion implantation process in which impurity ions are implanted using the mask pattern as an ion implantation mask, various annealing processes, and other processes. To generate a high manufacturing yield, these processes should be precisely managed and controlled according to a predetermined sequence using a semiconductor equipment control system. Most of these processes are controlled with respect to a single cassette in which a plurality of wafers in a single lot is mounted. For example, in a dry etching process using plasma, a cassette containing a lot of about twenty-five wafers is loaded into a load port of the semiconductor equipment to perform the dry etching process. The semiconductor equipment control system then reads a process program identification (ID) from the cassette loaded in the load port and searches for a corresponding process recipe (e.g., conditions for performing the process) pre-stored in the semiconductor equipment. The semiconductor equipment then performs the process according to the process recipe. FIG. 1 is a flowchart illustrating a conventional semiconductor equipment control method. Referring to FIG. 1, after loading a wafer lot (typically consisting of about 25 wafers) into the load port of the semiconductor equipment and searching for the process recipe corresponding to the lot, a host installed in the semiconductor equipment control system checks the entire recipe body of the searched process recipe (S10). Unfortunately, since the recipe body of the process recipe includes numerous specific conditions for performing a process (including, for example, a process environment, a process sequence, and a process type), it may take as long as ten minutes or longer to check the recipe body. To check the recipe body of the process recipe, the host compares values of the checked process recipe contents with the contents of a reference recipe (e.g., conditions appropriate to perform a process) stored in the host and then, determines whether the values of the checked process recipe are within an allowable tolerance range with respect to the reference recipe values (S30). When the values of the checked process recipe are within the allowable tolerance range with respect to the reference recipe values, the host transmits a predetermined control signal to the semiconductor equipment to allow the semiconductor equipment to perform a process according to the checked process recipe (S70). When the values of the checked process recipe are not within tolerance of the reference recipe, however, the host transmits an interlock signal to the semiconductor equipment to keep the semiconductor equipment from performing a process according to the process recipe (S50). Unfortunately, since the host searches for a process recipe and checks the recipe body of the process recipe whenever the lot is loaded in the semiconductor equipment, regardless of whether the process recipe has been modified, this unconditional checking procedure results in an unnecessary loss of process time. In particular, where the values of the pre-checked process recipes stored in the semiconductor equipment are within tolerance of the reference recipe, since most process recipes are not modified until they correspond to a subsequent lot, the conventional method of unconditionally checking the recipe body of every process recipe results in unnecessary delay and a corresponding decrease in productivity. The industry would therefore benefit from a system and method for controlling semiconductor equipment that result in increased productivity by reducing or eliminating the unnecessary delay associated with conventional unconditional recipe checking. According to various principles of the present invention, the problem described above is solved by providing a semiconductor equipment control system and method that is capable of minimizing the amount of time taken to check a recipe body of a process recipe. Accordingly, the principles of the present invention enable a semiconductor equipment control system and method that is capable of rapidly performing a semiconductor manufacturing process and thereby maximizing productivity. According to one aspect of the present invention, a semiconductor equipment control system may include semiconductor equipment having a process recipe stored therein. A host having a database in which a reference recipe is stored may be connected to the semiconductor equipment through a network. The host preferably compares a final modification time of the process recipe with a final modification time of the reference recipe to determine whether the recipe body checking process should be performed. When the final modification time of the process recipe is different from the final modification time of the reference recipe, the host may check the recipe body of the process recipe. When the final modification time of the process recipe is equal to the final modification time of the reference recipe, however, the semiconductor equipment preferably performs a process according to the process recipe. The recipe body may include a process environment, a process sequence, and a process type. To check the recipe body, the host may compare values of the process recipe and the reference recipe to determine whether the process recipe values are within an allowable tolerance range of the reference recipe values. When the values of the process recipe are within tolerance of the reference recipe, the host may update a final modification time of the reference recipe and perform a process according to the process recipe. However, when the values of the process recipe are not within tolerance of the reference recipe, the host may interlock the process. According to a still further aspect of the present invention, a semiconductor equipment control method may include checking a final modification time of a process recipe stored in the semiconductor equipment and comparing the checked final modification time of the process recipe with a final modification time of a reference recipe stored in a database. When the checked modification time of the process recipe is equal to the final modification time of the reference recipe, the control method may further include performing a process according to the process recipe. The control method may further include checking a recipe body of the process recipe when the checked modification time of the process recipe is different from the final modification time of the reference recipe. The recipe body may include a process environment, a process sequence, and a process type. The control method may also include comparing a value of the checked process recipe with a value of the reference recipe, and determining whether the value of the process recipe is within an established tolerance of the reference recipe. When the value of the process recipe is within tolerance of the reference recipe value, the control method may include updating a final modification time of the reference recipe. When the value of the process recipe is within tolerance of the reference recipe, the control method also preferably includes performing a process according to the process recipe. When the process recipe is not within tolerance of the reference recipe, the control method may include interlocking the process. In yet another embodiment, the semiconductor equipment may include a plurality of process chambers. In this embodiment, checking the final modification time of the process recipe may be performed separately with respect to each chamber installed in the semiconductor equipment. According to further principles of the present invention, a semiconductor equipment control method may include instructing a host to request a final modification time of a process recipe from semiconductor equipment connected to the host through a network, instructing the semiconductor equipment to provide the final modification time to the host, causing the host to compare the final modification time of the process recipe with a final modification time of a corresponding reference recipe stored in the host, and causing the host to transmit a control signal to the semiconductor equipment to perform a process according to the process recipe when the final modification time of the process recipe is equal to the final modification time of the reference recipe. When the final modification time of the process recipe is different from the final modification time of the reference recipe, the control method may include instructing the host to check a recipe body of the process recipe The recipe body may include a process environment, a process sequence, and a process type. The host may further compare values of the checked process recipe and the reference recipe, and determine whether the checked process recipe is within tolerance of the reference recipe. When values of the checked process recipe are within an allowable tolerance range of the values of the reference recipe, the host may control the semiconductor equipment to perform a process according to the process recipe and may further update the final modification time of the reference recipe stored therein to be equal to the final modification time of the process recipe. When the checked process recipe is not within tolerance of the reference recipe, the host may control the semiconductor equipment to interlock the process. The principles of the present invention will now be described more fully with respect to various preferred embodiments thereof. It should be recognized, however, that this invention may be embodied in many different forms and should not be construed as being limited to the specific embodiments set forth herein. Rather, these exemplary embodiments provide an enabling disclosure and best mode of practicing the invention that will convey the full scope of the invention to those skilled in the art. FIG. 2 is a schematic block diagram of a semiconductor equipment control system 100 in accordance with an exemplary embodiment of the present invention. Referring to FIG. 2, the semiconductor equipment control system 100 preferably includes a plurality of semiconductor equipment 130 for performing semiconductor manufacturing processes, a host computer 110 (“host”) for controlling the processes being performed by the semiconductor equipment 130, and a server 120 electrically connected to the semiconductor equipment 130 and the host 110 for bi-directionally transmitting various data and control signals. The semiconductor equipment 130 is preferably configured to perform various semiconductor device manufacturing processes including, for instance, a deposition process, a photolithography process, an etching process, an ion implantation process, various annealing processes, and other processes. The semiconductor equipment 130 preferably includes a data storage part, in which a process recipe 136 (containing conditions for performing a process), is stored in a file format. A controller for controlling operation of the semiconductor equipment 130 preferably operates the semiconductor equipment 130 to perform a process according to the contents of the process recipe 136 (referred to herein as the “recipe body”). The recipe body may include a process environment, a process sequence, and a process type for performing a process. In some instances, the recipe body may be partially modified by a process manager or the controller to finely adjust the process conditions and optimally perform a process. In such cases, time data corresponding to the time at which the modification was made (referred to herein as the “modification time”) of the process recipe 136 is automatically stored in the data storage part of the semiconductor equipment 130. The server 120 may be connected to the respective semiconductor equipment 130 and the host 110 through corresponding network connections. In addition, the server 120 is preferably configured to transmit various data and control signals from the host 110 to the semiconductor equipment 130, and to transmit data or other signals output from the semiconductor equipment 130 to the host 110. The semiconductor equipment 130, the server 120, and the host 110 may, for example, communicate with each other and share data using the Semiconductor Equipment Communication Standard (SECS) protocol. The SECS protocol provides a standard for enabling mutually recognizable communication between the semiconductor equipment 130, the server 120, and the host 110. The server 120 and the host 110 may also, however, communicate with each other and receive or transmit data through any generally known communication standard, such as the Transmission Control Protocol/Internet Protocol (TCP/IP), for instance. The host 110 is preferably configured to control a process being performed by the semiconductor equipment 130 through the server 120. In particular, the host 110 preferably includes a database 115, in which a reference recipe corresponding to a process being performed by the semiconductor equipment 130 is stored. The reference recipe may include conditions appropriate to perform the process, with specific values corresponding to those conditions. The reference recipe may further include an allowable tolerance range (including, for instance, upper and lower limit points) for each of the process conditions, within which tolerance range the process may appropriately be performed. The host 110 is preferably configured to determine under certain circumstances, whether the contents of the process recipe (the recipe body) is within tolerance of the reference recipe and to control whether or not the semiconductor equipment 130 performs the process according to that determination. Before checking the recipe body of the process recipe 136, however, the host 110 preferably compares a final modification time of the process recipe 136 with a final modification time of the reference recipe to determine whether checking the recipe body is even necessary. More specifically, the host 110 preferably controls the semiconductor equipment 130 to check the recipe body of the process recipe 136 only when the final modification time of the process recipe 136 is different from the final modification time of the reference recipe. When the final modification time of the process recipe 136 is equal to the final modification time of the reference recipe, the host 110 preferably performs the process in the semiconductor equipment 130 according to the process recipe 136 without checking the recipe body of the process recipe. Other functions of the host 110 will now be described in further detail with additional reference to FIG. 3, which illustrates a semiconductor equipment control method in accordance with an exemplary embodiment of the present invention. Specifically, FIG. 3 is a flowchart of a semiconductor equipment control method in accordance with an exemplary embodiment of the present invention. Referring to FIGS. 2 and 3, when a lot (typically consisting of about 25 wafers) is loaded into a load port of the semiconductor equipment 130 and after a process recipe 136 that corresponds to the lot is found, the host 110 of the semiconductor equipment control system 100 preferably checks a final modification time T1 of the process recipe 136 (S210). Checking the final modification time T1 of the process recipe 136 may be performed by instructing the host 110 to request the final modification time T1 of the process recipe 136 from the semiconductor equipment 130 through the network. The semiconductor equipment 130 may then be instructed to provide the final modification time T1 of the process recipe 136 to the host 110. The host 110 then compares the final modification time T1 of the process recipe 136 with a final modification time T2 of a corresponding reference recipe stored in a database 115 of the host to determine whether the final modification times T1 and T2 are equal to each other (S230). If the final modification time T1 of the process recipe 136 is equal to the final modification time T2 of the reference recipe, the host 110 then transmits a control signal to the semiconductor equipment 130 to perform a process according to the process recipe 136 (S270). On the other hand, if the final modification time T1 of the process recipe 136 is different from the final modification time T2 of the reference recipe, the host 110 checks the recipe body of the process recipe 136 (S240). Checking the recipe body of the process recipe 136 may be performed by instructing the host 110 to request data corresponding to the recipe body from the semiconductor equipment 130. The semiconductor equipment 130 may then be instructed to provide the recipe body information to the host 110. The host 110 then compares values of the process recipe 136 with values for the reference recipe conditions, and determines whether the values of the process recipe 136 being checked are within an allowable tolerance range of the reference recipe values (S250). If the process recipe 136 is within tolerance of the reference recipe, the host 110 preferably updates the final modification time T2 of the reference recipe stored in the host database 115 to be equal to the final modification time T1 of the process recipe 136 (S260). The host 110 may thereafter or simultaneously control the semiconductor equipment 130 to perform a process according to the process recipe 136 (S270). If, on the other hand, the value of the checked process recipe 136 is not within tolerance of the reference recipe, the host 110 may instruct the semiconductor equipment 130 to interlock the process (S280). As described above, this semiconductor equipment control method preferably only checks the contents of the recipe body of the process recipe against the contents of the recipe body of the reference recipe when the final modification time of the process recipe is different from the final modification time of the reference recipe. According to principles of the present invention, it is therefore possible to reduce time loss resulting from unconditionally checking the recipe body of the process recipe. FIG. 4 is a schematic block diagram of a semiconductor equipment control system 100′ in accordance with another exemplary embodiment of the present invention. The semiconductor equipment control system 100′ is similar to the semiconductor equipment control system 100 shown in FIG. 2 in many respects. A detailed description of those features previously described with respect to the embodiment shown in FIG. 2 will therefore be omitted. Referring to FIG. 4, the semiconductor equipment control system 100′ includes a plurality of pieces of semiconductor equipment 140. Each piece of semiconductor equipment 140 may include a plurality of process chambers 142 for performing the same process or different processes. Each piece of semiconductor equipment 140 also preferably includes a plurality of process recipes 146 corresponding to the respective chambers 142. Accordingly, a host 110 of the semiconductor equipment control system 100′ preferably compares a final modification time of each process recipe 146 with a final modification time of a corresponding reference recipe. The comparison is therefore performed with respect to each chamber 142 rather than with respect to each piece of equipment 140. The semiconductor equipment 140 can then be controlled using the control method described previously based on the result of each comparison. More particularly, when the process recipe 146 has the same final modification time as the corresponding reference recipe stored in the host 110, the respective process chamber 142 performs a process according to the process recipe 146 without checking the recipe body of the process recipe 146. On the other hand, when the process recipe 146 has a final modification time that is different from the corresponding reference recipe stored in the host 110, the corresponding process chamber 142 performs the process only if the process recipe body is within tolerance of the reference recipe body. The recipe body of the process recipe 146 must therefore be checked and compared to the reference recipe body using the host 110 (as described previously), only if the final modification times are different. Accordingly, even though the semiconductor equipment 140 includes a plurality of process chambers 142, the host 110 only checks the recipe body for a given process chamber 142 if the final modification time of the process recipe 146 corresponding to that process chamber 142 is different from a corresponding final modification time of a reference recipe. Using a semiconductor equipment control method in accordance with the principles of the present invention, it is therefore possible to more rapidly perform the semiconductor manufacturing process. As described above, a system according to principles of the present invention preferably compares the final modification times of the process recipe and the reference recipe to determine whether or not to check the recipe body of the process recipe. In other words, each recipe version may be used to determine whether or not to check the recipe body of the process recipe, or a determination symbol may be added to determine whether a separate modification has been made to each recipe. According to principles of the present invention, since a recipe body of a process recipe is checked only when the process recipe stored in semiconductor equipment has been modified by a manager or a controller, it is possible to appropriately perform a process while minimizing an amount of time taken to check the recipe body of the process recipe. It is therefore possible to more rapidly perform the semiconductor manufacturing process to maximize productivity. While this invention has been described in connection with what are presently considered to be the most practical and preferred embodiments, it is to be understood that the invention is not limited to the disclosed embodiments, but is instead intended to cover various modifications thereto within the spirit and the scope of the invention as set forth in the following claims. |
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048790893 | abstract | In a liquid metal cooled fast breeder nuclear reactor of the pool type, the hot and cold pools (13, 19) are separated by an intermediate plenum (24) which, under steady state conditions, encloses a volume of substantially stagnant liquid metal to form a thermal barrier between the hot and cold pools (13, 19). To avoid undesirably large temperature differentials developing for example in the event of large temperature excursions within the hot pool (13) as a result of a reactor trip, the intermediate plenum (24) is designed to permit, in such circumstances, rapid interchange of liquid metal coolant between the hot pool and the intermediate plenum, for example by way of a thermal siphon arrangement (30). |
description | This invention was developed under Contract No. DE-NA0003525 awarded by the United States Department of Energy/National Nuclear Security Administration. The Government has certain rights in this invention. The application generally relates to a system and method for a non-contact rapid reader system for reflective particle tags, or labels, (RPT) applied to containers and other articles for monitoring. Containment and surveillance measures are critical to any verification regime in order to monitor certain highly secured and restricted activities, e.g., transportation of nuclear fuel and its components across international borders; to detect undeclared activities related to national security and restricted activities; to verify the integrity of equipment or items; to reduce inspector burden; and to maintain a chain of custody between inspections. A tag is an exemplary measure used to establish the identity of an accountable item and maintain the chain of custody for the respective item. Tags must also provide evidence of tampering of the tag itself, e.g., counterfeiting or substitution, and if applied in an appropriate manner, e.g., across a seam of a container, a tag may also provide evidence of tampering with the item. Continual improvement of measures such as tags is required to counteract technical advances of adversaries which could render C/S equipment obsolete with a single technical advancement. Furthermore, new architectures are required to respond to changing requirements arising from the introduction of new procedures or approaches, and it is often desirable to incorporate technological advances that provide efficiency gains or allow deployment in new application spaces. The RPT was developed to identify items that must be accounted for under international treaties. In most instances the tag, or RPT, is composed of an article with unique optical characteristics, e.g., specular hematite particles randomly dispersed in a clear, adhesive polymer matrix. Reflective particle tags (RPT) derive their unique identities through utilization of thousands of microscopic reflective elements randomly suspended in a clear adhesive matrix. For verification of the authenticity, an illumination/imaging system is used to “read” information about precise positions and orientations of faceted particles. SNL developed the original Reflective Particle Tag (RPT) system, comprising a tag and an imager, in the 1990's to identify treaty-accountable items. Since then, the RPT system has evolved with advances in computing, imaging, and materials, and is considered a robust, low-cost, hard-to-counterfeit passive tagging system for treaty verification. However, a limitation of the current system is the need to mechanically dock the reader with the tag, which prevents its use in many situations. This paper discusses R&D at SNL to develop a non-contact handheld imaging system that will allow RPT system use in new scenarios and allows automation. The RPT architecture is effective for detection of counterfeiting and removal of tags. Furthermore, RPTs require no power source, and maintain stability through temperature extremes, rough handling, and years of service. Such attributes make RPTs suitable for applications with strict facility acceptance requirements and for deployments in which a semi-permanent tag should be attached to an item to be monitored. However, the current RPT system referred to as a contact-type RPT system, suffers from certain deficiencies that limit potential applications. The contact-type RPT derives its security capability through precise alignment between a reader and the RPT, and relies on tightly collimated illumination beams and a small aperture to allow only facets oriented within approximately one degree of the optimal direction to contribute to the recorded image. In order to achieve such precision, the reader must be placed in contact with the flat frame in which the RPT is mounted or attached. However, such physical contact may be undesirable or not permitted by a facility owner. In addition, the use of a flat frame is incompatible where tags must be affixed to complex geometries and curved surfaces. What is needed is a system and/or method for rapidly testing and evaluating optical designs of computational and compressive imaging sensing (CS) systems. The term compressive sensing (CS) is used interchangeably with computational imaging, for purposes of this disclosure. One embodiment relates to a reflective particle tag reader system including a read head assembly. The read head assembly has a camera, illuminators, and a rigid frame portion for supporting the camera and the illuminators. The illuminators are mounted to the frame and directed to illuminate a focal point located opposite the camera. The focal point is where a reflective particle tag is place. A computer having a display, a processor, a data communication input and output means, and a data storage device is in data communication with the camera to receive and store images of the reflective particle tag that are acquired by the camera. The computer is programmed to process video images and to quantify a positional alignment parameter and an angular alignment parameter of the reader with respect to the reflective particle tag. A rapid burst of image frames is obtained in response to the positional alignment and the angular alignment parameters being within a predetermined tolerance and identity of the reflective tag is established between a first image set acting as an authenticating reference and a second image set used for verification. Another embodiment relates to a read head assembly having a camera, illuminators, and a rigid frame portion for supporting the camera and the illuminators. The illuminators are mounted to the frame and directed to illuminate a focal point located opposite the camera. Yet another embodiment discloses a non-transitory computer-readable storage medium having stored thereon instructions which, when executed by one or more processing units, cause the one or more processing units to perform a method for maintaining and verifying authenticity of a reflective particle tag. The method includes illuminating a focal point located opposite the camera, placing a reflective particle tag at the focal point, receiving by a computer in data communication with a camera one or more images of the reflective particle tag acquired by the camera, processing video images and quantifying a positional alignment parameter and an angular alignment parameter of the reader with respect to the reflective particle tag, obtaining a rapid burst of image frames in response to the positional alignment and the angular alignment parameters being within a predetermined tolerance; and authenticating an identity of the reflective tag between a first image set and a second image set. The disclosed invention overcomes a number of deficiencies present in a contact-type RPT system by providing a handheld tag reader and associated computational hardware capable of identifying an RPT without coming into physical contact with the RPT. In addition, the disclosed RPT reader system is compatible for reading tags attached to items having complex geometries, such as curves or ridges, and reduces the time required for inspection in harsh or environmentally restricted locations. The RPT reader system further provides the capability to automate repetitive tasks, e.g., reading RPTs attached to cylinders with uranium hexafluoride (UF6) for use in nuclear fuel processing and enrichment. The disclosed handheld reader system combines many characteristics of the contact RPT system such as utilization of multiple illumination angles to record unique images of only those particles whose orientations match the reflection criterion. However, since a non-contact handheld system may be compromised by motion and misalignment introduced by the equipment operator, the new system reduces reliance on precise registration of the imager with the tag. While it is desirable to provide the greatest precision possible in terms of alignment, an advantage of the disclosed RPT reader system is the ability to recognize when an acceptable alignment condition has been momentarily achieved. Another advantage of the RPT reader system is resistance to counterfeiting and removal without detection. The handheld RPT reader includes an optical system and vision processing software allows the acquisition of high quality images without direct contact with the RPT. The disclosed embodiments rely on a handheld read head in data communication with a processing device, e.g., a desktop computer or tablet computer, having special purpose circuit boards. Alternative exemplary embodiments relate to other features and combinations of features as may be generally recited in the claims. Before turning to the figures which illustrate the exemplary embodiments in detail, it should be understood that the application is not limited to the details or methodology set forth in the following description or illustrated in the figures. It should also be understood that the phraseology and terminology employed herein is for the purpose of description only and should not be regarded as limiting. Referring to FIG. 1, a scanning electron microscope image of the faceted specular hematite particles 12 used in the RPT system 100 (FIG. 4). FIG. 1 shows a magnified view of a commonly used RPT 10 including hematite particles 12 with dimensions approximately 80 μm with flat, reflective facets. FIG. 2 shows an exploded view for an exemplary RPT assembly 24 for a prior art contact-type RPT. A cover portion 22 is placed over a docking frame 26 to protect RPT 10 from damage. A tag frame 28 supports RPT 10 in aperture and attaches to docking frame 26. Docking frame 26 rigidly supports a reader for alignment with RPT when reading patterns associated with RPT. Frames 28 held the read head securely in place since the data processing was slow, and “on the fly” image capture was previously not possible. In a contact-type prior art RPT system, in order to inspect tag 10, reader 100 is physically attached to a frame 18 for precise alignment and records images using each of the four illuminators. For each of the illumination angles, only a subset of the hematite facets will be oriented in such a manner as to redirect the incident illumination beam toward the aperture of the camera, collectively referred to as the reflection criterion. This subset will appear as small bright spots in a recorded image 40 as shown in FIG. 3. In this manner, a sequence of four complex and highly unique patterns 46 may be recorded and stored in a data storage device (not shown). The recorded patterns 46 may be used to physically authenticate the tag at a later date. Optionally, a unique, barcode-like identifier (ID) 42 may be placed at a center line 44 of RPT 10 to identify RPT 10 (FIG. 3) and allow rapid retrieval of reference images. Once a tag is set, an inspector can return to the item, attach the reader, compare barcode or other IDs 42, then reflective patterns 46 in order to determine if tag 10 patterns 46 are a match. As shown in FIG. 4, an RPT reader 100 includes a camera 14 and multiple collimated illuminators 16a, 16b, 16c, arranged at different angles with respect to RPT 10. While three illuminators are shown in the exemplary embodiment, more or fewer illuminators may be used as desired for the application. Likewise, other parametric values used in the embodiments described herein are provided by way of example and not limitation. In another embodiment, a color charge-coupled device (CCD) and three LEDs that are red, green, and blue may be used as illuminators. In this embodiment, the system can capture images from the three LEDs simultaneously to increase system speed. A schematic diagram of the read head 50 of the RPT system 100 is shown in FIG. 4. A read head 50 includes a rigid, preferably light-weight circular frame portion 18 with handles 32. Three high-power LED illuminators 16a, 16b, 16c, are mounted to the frame and aimed toward a point approximately 10 cm below the center of circular frame portion 18. In contrast to the tightly collimated illumination beams of the contact RPT system, the illuminators of the new system project approximately F/2 beams—representing a ratio of the system's focal length to the diameter of the aperture—containing a larger range of illumination angles. This is achieved using collimation optics and diffusers. The LEDs of illuminators 16 emit light in a relatively narrow spectral band so that can be filtered by a narrow bandpass filter (not shown) to reject most of the ambient illumination. The illuminators 16 may be individually or simultaneously powered. In one embodiment, illuminators 16 may be rapidly strobed sequentially. In one embodiment, the bandpass filter may be a flat plate of colored glass mounted in front of the camera lens, e.g., if the illuminators 16 produce light at a wavelength λ=440 nm, a filter passing 440 nm+/−10 nm will block >95% of the visible light. Computer Vision System: For obtaining near optimum alignment of reader 100 with RPT 10, a high frame-rate computer vision processing system (not shown) quantifies positional and angular alignment errors of the reader with respect to RPT 10. With each frame, a set of fiducials is projected on the display screen to assist the operator or inspector to improve the alignment of reader 100. When the alignment satisfies a predetermined criterion, a rapid burst of image frames is obtained to be used for verification of the image sets. Thus, the computer vision system essentially comprises two major components—alignment and verification. In one embodiment, the rapid burst of camera images is obtained while in synchronization with the strobing of the illuminators. FIG. 5 illustrates a view from the computer vision system of an RPT according to an embodiment of the disclosure. Referring next to FIG. 5, the computer vision system relies on image features and focus measures that are recorded at the time an RPT 10 is attached to an item for tracking. For each RPT, a set of features, or data set, in particular features calculated using the Speeded Up Robust Features, or SURF, algorithm, their descriptors and focus measures is stored on RPT system 100 computer. The SURF algorithm is well known in the art and provides a local feature detector and descriptor. SURF is used for tasks such as object recognition and image registration. Its feature descriptor is based on the sum of the Haar wavelet response around the point of interest. During alignment, the vision system analyzes each new frame, or image, recorded by camera 14 and rapidly acquires a new set of image features for the incoming image. The acquired features are compared to the stored features, and, using only features that provide a good match between the current and stored feature sets, a homography matrix is calculated which provides the lateral displacement and azimuthal rotation between the acquired features and stored features. A pair of crosshair fiducials 60, 62 (FIG. 6) is projected on the computer screen, with one crosshair 60 representing the coordinate system of an imager chip and the other crosshair 62 representing the coordinate system of the stored feature set. When the fiducials 60, 62 are aligned, three of six possible degrees of orientation and rotation are aligned. Degrees of freedom include the lateral position in the plane of the tag and the in-plane rotation. The remaining degrees of freedom are the 2-D tilt of the plane and the focus. The remaining three degrees of freedom are aligned using the currently acquired set and the corresponding stored set of focus measures. Each of these focus measures is obtained at a different spatial location within the image 64. If the reader 100 is angled relative to the conditions that were used for the stored data, then better focus measures will be obtained for some regions of the image 64 than other regions. Other parts of image 64 are sharply imaged on frames that are near in time. A balance bubble fiducial may be projected on the computer display to indicate relative balance of all the focus measures. When the cross-hair fiducials are matched and the balance bubble is centered, then all degrees of freedom have been aligned. FIG. 5 shows a screenshot of the computer display with the projected fiducials for the tag reader application during the alignment phase. Red and green crosshairs are used in one embodiment to distinguish from one another. Green crosshairs 60 are aligned with the imager and red crosshairs 62 are aligned with the tag. Red and green circles 68, 70 indicate the focus and tilt errors. When the crosshairs 60, 62 overlap within a pre-specified error tolerance, and circles 68, 70 are centered at the origin, the system acquires a rapid burst of high resolution images that can be used for verification. Alternatively signals or schemes other than fiducials may be used to characterize the quality of the alignment and focus for verification purposes. FIG. 6 shows a set of fiducials overlaying a blank image for contrast. FIG. 7 is a display screen for adjusting and displaying fiducial parameters, settings and controls for the operator. In practice it may not be possible for an operator to perfectly align the fiducials 60, 62, so a predetermined set of tolerances is utilized to determine whether the alignment of fiducials 60, 62 is sufficiently close to allow image acquisition and verification of the tag. Using binned images to reduce the data set size eightfold, the system software can complete the image analysis and fiducial projection at a rate of 25 frames per second. During this alignment phase, all three illuminators 16a, 16b, 16c, may be constantly illuminated. Once the computer vision system deems that the alignment is appropriate, the system acquires a burst of ˜100 full resolution (2048×2048) frames at a rate of 90 frames per second. In one embodiment, during the burst acquisition, illuminators 16 are sequentially strobed to allow acquisition of frames using each illumination condition. Preferably, to obtain maximum acquisition speed for frames, vision processing tasks are performed after the burst occurs. After the burst acquisition, the image sequence is analyzed based, e.g., the best focus measures and the top four images per illumination condition are retained for a total of twelve images per acquisition. These high-quality frames are saved and are further processed using another SURF algorithm for verification processing. The ability to align the reader with the tag and acquire images for verification is sufficient for tag verification; however, full resolution images provide a more rigorous verification procedure where increased confidence may be desired. The security of an RPT system 100 depends upon a number of factors including: the number of particles appearing in each image; the angular tolerance for satisfying the reflection criterion (i.e., how much of the light is specularly reflected from a facet and collected by the camera lens); the positional tolerance for locating the centroid of a facet; and the amount of spatial information related to the shapes and sizes of the facets. The handheld system differs in several important aspects from that of the contact-type RPT system. It is not possible to determine an absolute relative comparison value for the handheld vs. non-contact system security, since a variable element of system security depends upon how difficult it would be for an adversary to replicate an existing tag. An exemplary estimate of security may be based on the number of particles within the images, their angular and positional tolerances, and a shape information factor. To do this, a security figure of merit representing confidence in the uniqueness of a tag image is defined as: S = N · P ( Δ θ ) S · ( Δ x ) 2 Equation 1 where N is the number of particles appearing in the images, P is the number of degrees of freedom describing the shape of the particle, AO is the measurement tolerance of the particle's tilt angles, and Δx is the measurement tolerance of the particle's centroid. It is assumed that the camera images, e.g., a 15 mm×15 mm field of view. Also, assume the tag utilizes hematite particles with an average size of ˜80 micrometers (μm). Other parameters of the two RPT systems for estimated values are shown in table 1: TABLE 1ContactNon-Contact RPT SystemHandheld RPT SystemImager size1.3MP4MPObject space f-number8.45.6Number of illuminators43Illuminator f-number202Resolvable spot diameter19μm12μm(referred to tag plane) The average number of particles that will appear in each image depends upon the f-numbers of the illuminators and the camera lens, and is greater by a factor of ˜12 for the handheld RPT system due to its faster optics. However, the increased average number of particles carries with it a factor of ˜3.5 loosening of the measurement tolerance of the particle's tilt angle which is Δθ≈±1.7 degrees for the contact RPT system and Δθ≈±6 degrees in the handheld system. Note that since the number of particles visible in the images scales as the square of the angular tolerance factor, these factors will cancel in the comparison of the security figures of merit. The higher spatial resolution of the non-contact handheld system directly leads to a larger number of degrees of freedom describing the shapes of the particles. To estimate this effect, an average particle patch of 300 μm is assumed in order to account for irregularities in particle shape. Using resolvable spot diameters for the two systems presented in the table, we obtain the number of resolution elements along the perimeter as P=25 for the non-contact handheld reader, and P=16 for the contact RPT. In a similar fashion, the improved spatial resolution of the non-contact handheld reader allows the particle centroids to be determined more precisely. This is partly due to the smaller resolution element size and partly due to the larger number of resolution elements appearing on the surface (perimeters of the particles). The centroid positions can be estimated to an accuracy of Δx≈±3.4 μm for the handheld system and Δx≈±6.7 μm for the contact RPT system. Assembling these estimates yields a security figure of merit, S, for the non-contact type handheld reader is greater than that of the contact RPT by a factor of three. Although the exemplary embodiments disclosed herein refer to hematite as the reflective particle tag, any crystalline material that tends to fracture so as to leave flat faces is possible. The refractive index of the crystal is preferably significantly different than that of the plastic (n˜1.5) in which the crystals are captured. Also, metallic crystal material, e.g., pyrite, as well as dielectric materials like diamond, sapphire, Graphite, Ge, Si, and rare materials, Ho—Mg—Zn, may be substituted for hematite. Facet sizes are preferably 20 microns <D<˜200 microns. While the exemplary embodiments illustrated in the figures and described herein are presently preferred, it should be understood that these embodiments are offered by way of example only. Accordingly, the present application is not limited to a particular embodiment, but extends to various modifications that nevertheless fall within the scope of the appended claims. The order or sequence of any processes or method steps may be varied or re-sequenced according to alternative embodiments. The present application contemplates methods, systems and program products on any machine-readable media for accomplishing its operations. The embodiments of the present application may be implemented using an existing computer processor, or by a special purpose computer processor for an appropriate system, incorporated for this or another purpose or by a hardwired system. It is important to note that the construction and arrangement of the non-contact reflective particle tag reader system as shown in the various exemplary embodiments is illustrative only. Although only a few embodiments have been described in detail in this disclosure, those skilled in the art who review this disclosure will readily appreciate that many modifications are possible (e.g., variations in sizes, dimensions, structures, shapes and proportions of the various elements, values of parameters, mounting arrangements, use of materials, colors, orientations, etc.) without materially departing from the novel teachings and advantages of the subject matter recited in the claims. For example, elements shown as integrally formed may be constructed of multiple parts or elements, the position of elements may be reversed or otherwise varied, and the nature or number of discrete elements or positions may be altered or varied. Accordingly, all such modifications are intended to be included within the scope of the present application. The order or sequence of any process or method steps may be varied or re-sequenced according to alternative embodiments. In the claims, any means-plus-function clause is intended to cover the structures described herein as performing the recited function and not only structural equivalents but also equivalent structures. Other substitutions, modifications, changes and omissions may be made in the design, operating conditions and arrangement of the exemplary embodiments without departing from the scope of the present application. As noted above, embodiments within the scope of the present application include program products comprising machine-readable media for carrying or having machine-executable instructions or data structures stored thereon. Such machine-readable media can be any available media which can be accessed by a general purpose or special purpose computer or other machine with a processor. By way of example, such non-transitory machine-readable media can comprise RAM, ROM, EPROM, EEPROM, CD-ROM or other optical disk storage, magnetic disk storage or other magnetic storage devices, or any other non-transitory computer-readable medium which can be used to carry or store desired program code in the form of machine-executable instructions or data structures and which can be accessed by a general purpose or special purpose computer or other machine with a processor. When information is transferred or provided over a network or another communications connection (either hardwired, wireless, or a combination of hardwired or wireless) to a machine, the machine properly views the connection as a machine-readable medium. Thus, any such connection is properly termed a non-transitory machine-readable medium. Combinations of the above are also included within the scope of machine-readable media. Machine-executable instructions comprise, for example, instructions and data which cause a general purpose computer, special purpose computer, or special purpose processing machines to perform a certain function or group of functions. It should be noted that although the figures herein may show a specific order of method steps, it is understood that the order of these steps may differ from what is depicted. Also, two or more steps may be performed concurrently or with partial concurrence. Such variation will depend on the software and hardware systems chosen and on designer choice. It is understood that all such variations are within the scope of the application. Likewise, software implementations could be accomplished with standard programming techniques with rule based logic and other logic to accomplish the various connection steps, processing steps, comparison steps and decision steps. |
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abstract | In a decoding method of maximum a posteriori probability, (1) backward probabilities are calculated in a reverse direction from an Nth backward probability to a first backward probability, an math backward probability, m(sxe2x88x921)th backward probability, . . . , m2th backward probability are saved discretely, an m1th backward probability to the first backward probability are saved continuously, first forward probability is calculated, a first decoded result is obtained using the first forward probability and the saved first backward probability, and second to m1th decoded results are obtained in similar fashion. (2) Thereafter, backward probabilities up to an (m1+1)th backward probability are calculated and saved starting from the saved m2th backward probability, (m1+1)th forward probability is calculated, an (m1+1)th decoded result is obtained using the (m1+1)th forward probability and the saved (m1+1)th backward probability, (m1+2)th to m2th decoded results are obtained in similar fashion and (3) (m2+1)th to Nth decoded results are subsequently obtained in similar fashion. |
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046541861 | summary | FIELD OF THE INVENTION The present invention relates to the determination of the power of a pressurized water nuclear reactor which comprises at least one cooling loop and preferably three or four loops. It is important to know the power of a nuclear reactor at any time, particularly for safety reasons. PRIOR ART The present practice is to employ neutron detectors installed outside the reactor vessel for measuring the power. Such detectors give signals with a fast response but have the disadvantage of giving signals of low precision, particularly during periods of transient operation. To obtain more precise results, the reactor power can be computed from a heat balance calculated from measurements of the temperature of the primary fluid in the cold branch and in the hot branch of the primary loops. However, while the results obtained in this way are precise, they are nevertheless slow because of the high time constant of the temperature variations. Devices which permit the power of a nuclear reactor to be correctly determined in respect of both precision and speed are known, e.g. from, French Pat. No. 2,373,057, which describes an apparatus permitting the reactor power to be determined from the increase in core enthalpy, this enthalpy increase being computed by making use of the velocity of sound in the fluid in the hot and cold branches of the primary circuit. However, this apparatus involves installation of sensors on the primary circuit pipework, the major disadvantage being the increase in the number of connections to the said pipework and the resulting increase in the difficulty of lagging it. U.S. Pat. No. 3,752,735 describes a device which makes it possible to produce a signal representing the thermal power of the core by using the measurement of hot branch temperature and cold branch temperature and compensating these temperature measurements dynamically, following a formula which involves the time derivative of the difference between the hot branch temperature and the cold branch temperature. Such a device permits a fast response signal to be obtained, but the formula which is employed to obtain this signal is only approximate and renders the signal somewhat inaccurate. French Patent Application No. 2,416,531 describes a process for determining the power of a nuclear reactor in which a thermal power signal which has a high time constant but is relatively precise is combined with a neutron power signal which has a fast response but is less precise. Nevertheless this process does not produce very precise results in a transient regime, because it merely adjusts the thermal power signal by means of a neutron power signal which is delayed by a gain control unit 6 and an integrating unit 7 (FIG. 1). It does not take into account numerous data which can vary in a transient regime. Furthermore, to date there has been no known simple device enabling both the primary power and the secondary power of a nuclear reactor to be obtained rapidly and precisely both in steady state operation and in transient operation. SUMMARY OF THE INVENTION The object of the present invention is to remedy the disadvantages of the aforesaid processes and devices. It relates to a device for fast and precise determination of the power of a pressurized water nuclear reactor having one or more cooling loops, in steady state operation and during periods of transient operation. Moreover, in a preferred embodiment, it also makes it possible to obtain, in an equally simple, fast and precise manner, the secondary power, i.e., the power of the steam generators in each cooling loop of the reactor. According to the invention the device comprises, for each of its cooling loops: means for measuring, on the one hand, the neutron flux and, on the other hand, the temperature of the primary fluid at a point in the cold branch and at a point in the hot branch, means for computing the enthalpy of the primary fluid in the cold branch and in the hot branch from the said temperature measurements, a register which computes the enthalpy increase of the primary fluid as it crosses the core, from the difference between the enthalpy in the hot branch and the enthalpy in the cold branch delayed by a time shift operator expressing the average time of transit of a molecule of fluid between the temperature measurements points in the cold branch and in the hot branch, a multiplier of the increase in the enthalpy by the flow rate of the primary fluid, a comparator of the thermal power signal which is obtained and of the neutron power signal which is measured and made dynamically equivalent to the thermal power signal, and a corrector of the neutron power measurement signal depending on the signal produced by the said comparator. In a preferred embodiment of the invention, the neutron power measurement signal is made dynamically equivalent to the thermal power signal by means of a point model of heat transfer between the nuclear flux and the thermal flux to the primary fluid, of a point model of heat transfer of the fluid in the core corresponding to the time of transit of a molecule of primary fluid from the center of the core, to the outlet of the core and a time shift operator expressing the time of transit of a molecule of primary fluid from the core outlet to the point of measurement of the hot branch temperature. The corrector of the neutron power measurement signal is preferably an integrator. Moreover, it is preferable to correct the neutron power measurement signal for variations of the temperature measured in the cold branch before this signal is compared to the thermal power signal. It is advantageous to supplement the device according to the invention with a filter of the primary flow rate signal which is situated upstream of the said multiplier in order to allow for variation of the average time of transit of a molecule of primary fluid from the cold branch to the hot branch depending on the primary flow rate. It is further preferred to supplement the device for fast and precise determination of the primary power of a nuclear reactor as described above, by a device for rapid and precise determination of the secondary power of the reactor, comprising, for each loop, in addition to the aforesaid means for measuring temperature and for computing the enthalpy in the cold branch and in the hot branch, imprecise but fast means for determining the thermal power produced by the steam generator associated with the reactor from measurements of the temperature and the flow rate of the feed water and of the pressure and the flow rate of the steam from the steam generator, a register computing the reduction in the enthalpy of the primary fluid as it crosses the steam generator from the difference between the enthalpy in the hot branch and in the cold branch, the enthalpy in the hot branch being delayed by a time shift operator expressing the time of transit of a molecule of fluid between the two temperature measurement points, a multiplier of the decrease of enthalpy by the flow rate of the primary fluid, a comparator of the signal, which is obtained at the multiplier output, of the thermal power absorbed by the steam generator to the said signal, as determined above, of the thermal power produced by the steam generator, this signal being made dynamically equivalent to the signal of thermal power absorbed by the steam generator, and a corrector of the signal of power produced by the steam generator depending on the signal produced by the said comparator. Preferably, the measurement signal of the thermal power produced by the steam generator is made dynamically equivalent to the signal of the thermal power absorbed by the steam generator by means of a point model of heat transfer between the secondary fluid and the primary fluid, a point model of heat transfer of the primary fluid in the steam generator corresponding to the time of transit of a molecule of primary fluid from the center of the steam generator to the outlet of the steam generator, and a time shift operator expressing the time of transit of a molecule of primary fluid from the outlet of the steam generator to the point of measurement of the cold branch temperature. It is preferred to use an integrator to correct the measurement signal of the thermal power which is absorbed by the steam generator. The dynamics of the measurements of the primary fluid temperature in the cold branch and the hot branch are furthermore compensated by a phase lead corrector which is placed at the output of the multiplier of the variation of the enthalpy of the primary fluid by the flow rate of the primary fluid . |
053295692 | abstract | A composite window structure is described for transmitting x-ray radiation and for shielding radiation generated debris. In particular, separate layers of different x-ray transmissive materials are laminated together to form a high strength, x-ray transmissive debris shield which is particularly suited for use in high energy fluences. In one embodiment, the composite window comprises alternating layers of beryllium and a thermoset polymer. |
description | This application claims the benefit of PCT/CN2003/000986 filed Nov. 20, 2003. The present invention relates to a collimator, as well as an X-ray irradiator and an X-ray apparatus. Particularly, the invention is concerned with a collimator for restricting an irradiation range of X-ray, as well as an X-ray irradiator and an X-ray apparatus both provided with such a collimator. In an X-ray irradiator there is used a collimator for restricting an irradiation range of X-ray. The collimator has an aperture permitting X-ray to pass therethrough and has a structure such that X-ray cannot pass through the collimator except the aperture. With this structure, the irradiation range of X-ray can be adjusted. A collimator having a variable aperture is provided with movable plate members, that is, blades having X-ray absorbability. As the blades there are used a pair of blades opposed to each other at respective end faces. The pair of blades are movable in directions opposite to each other in a plane parallel to their surfaces. For expanding the aperture, the pair of blades are moved in directions away from each other, while for narrowing the aperture, the blades are moved toward each other. As a collimator developed for reducing the size thereof without sacrificing a variable range of an aperture there is known one in which blades are constructed of a flexible material and are wound round drums and, for expanding the aperture, are wound inwards round the drums, while for narrowing the aperture, they are paid out from the drums (see, for example, Patent Literature 1). [Patent Literature] Japanese Published Unexamined Patent Application No. 2002-355242 (pages 2 to 3, FIGS. 1 to 2) The above known collimator involves the problem that it is necessary to use a special material flexible and superior in X-ray absorbability as the blase material. Therefore, it is an object of the present invention to provide a collimator which can be reduced in size without using any special material and without sacrificing an aperture, as well as an X-ray irradiator and an X-ray apparatus both provided with such a collimator. (1) The present invention, in one aspect thereof for solving the above-mentioned problem, resides in a collimator comprising: a pair of first plate members each having X-ray absorbability, movable in a direction parallel to a surface thereof, and defining an X-ray passing aperture by a spacing between respective end faces opposed to each other; and a pair of second plate members each X-ray absorbability, the second plate members, in order to block other X-rays than the X-ray passing through the aperture, being connected at respective one ends by hinges to end portions of the pair of first plate members opposite to the mutually opposed end faces of the first plate members and being supported at respective opposite ends so as to be movable obliquely with respect to the moving direction of the first plate members with movement of the first plate members. (2) The present invention, in another aspect thereof for solving the above-mentioned problem, resides in an X-ray irradiator comprising: an X-ray tube; and a collimator for collimating X-ray generated from the X-ray tube. The collimator comprises: a pair of first plate members each having X-ray absorbability, movable in a direction parallel to a surface thereof, and defining an X-ray passing aperture by a spacing between respective end faces opposed to each other; and a pair of second plate members each having X-ray absorbability, the second plate members, in order to block other X-rays than the X-ray passing through the aperture, being connected at respective one ends by hinges to end portions of the pair of first plate members opposite to the mutually opposed end faces of the first plate members and being supported at respective opposite ends so as to be movable obliquely with respect to the moving direction of the first plate members with movement of the first plate members. (3) The present invention, in a further aspect thereof for solving the above-mentioned problem, resides in an X-ray apparatus comprising: an X-ray tube; a collimator for collimating X-ray generated from the X-ray tube and applying the collimated X-ray to an object to be radiographed; and a detector means for detecting the X-ray which has passed through the object to be radiographed. The collimator comprises: a pair of first plate members each having X-ray absorbability, movable in a direction parallel to a surface thereof, and defining an X-ray passing aperture by a spacing between respective end faces opposed to each other; and a pair of second plate members each having X-ray absorbability, the second plate members, in order to block other X-rays than the X-ray passing through the aperture, being connected at respective one ends by hinges to end portions of the pair of first plate members opposite to the mutually opposed end faces of the first plate members and being supported at respective opposite ends so as to be movable obliquely with respect to the moving direction of the first plate members with movement of the first plate members. In the invention in each of the above aspects, the second plate members which constitute blades together with the first plate members are bent from connections and are movable obliquely, so that an external form of the collimator can be made small without using any special material such as a flexible X-ray absorbing material and without sacrificing the aperture. From the standpoint of simplifying the support structure it is preferable that the opposite ends of the second plate members be supported by guide grooves formed obliquely relative to the moving direction of the first plate members and pins engaged in the guide grooves. Likewise, from the standpoint of enhancing the freedom of setting an irradiation range in the moving direction of the first plate members it is desirable that the pair of first plate members be movable independently of each other. According to the present invention, it is possible to provide a collimator which permits the reduction of size, as well as an X-ray irradiator and an X-ray apparatus both provided with such a collimator, without using any special material and without sacrificing the aperture. Further objects and advantages of the present invention will be apparent from the following description of the preferred embodiments of the invention as illustrated in the accompanying drawings. An embodiment of the present invention will be described in detail hereinunder with reference to the drawings. FIG. 1 illustrates an X-ray apparatus schematically. This apparatus is an example of a mode for carrying out the invention. With the construction of this apparatus there is shown an example of a mode for carrying the invention with respect to the apparatus thereof. In this X-ray apparatus, as shown in the same figure, X-ray generated from an X-ray tube 1 is diaphragmed by an X-ray diaphragm 3 and is collimated by a collimating plate 500 disposed within a collimator 5, then the collimated X-ray is applied toward an object 7 to be radiographed and transmitted X-ray is detected by a detector 9. The X-ray tube 1 is an example of a mode for carrying out the invention with respect to the X-ray tube defined herein. The collimator 5 is an example of a mode for carrying out the invention with respect to the collimator of the invention. The detector 9 is an example of a mode for carrying out the invention with respect to the detector means defined herein. The portion comprising the X-ray tube 1, X-ray diaphragm 3 and collimator 5 is an example of a mode for carrying out the invention with respect to the X-ray apparatus of the invention. With the construction of this apparatus there is shown an example of a mode for carrying out the invention with respect to the X-ray apparatus of the invention. The collimator 5 is an example of a mode for carrying out the invention with respect to the collimator of the invention. With the construction of this apparatus there is shown an example of a mode for carrying out the invention with respect to the collimator of the invention. The X-ray tube 1 has an anode 101 and a cathode 103, and X-ray is generated from a point of collision (focus) of electrons which are emitted from the cathode 103 toward the anode 101. The X-ray thus generated is applied to the object through the X-ray diaphragm 3 and the collimator 5. The X-ray diaphragm 3 is constructed of an X-ray absorbing material such as lead for example. The collimating plate 500 in the collimator 5 is also constructed of an X-ray absorbing material such as lead for example. The X-ray diaphragm 3 shapes the X-ray generated from the X-ray tube 1 so that the X-ray becomes a quadrangular pyramid-like beam with an X-ray focus on the anode 101 as a vertex. The collimator 5 defines an X-ray irradiation field V by an aperture which is formed by the collimating plate 500. The aperture is variable to adjust the X-ray irradiation field V. Reference will be made to the collimating plate 500 in the collimator 5. FIG. 2 shows the construction of a principal portion of the collimating plate 500. In the same figure, three mutually perpendicular directions are assumed to be x, y, and z directions, z being the vertical direction. The X-ray is radiated from above. As shown in the same figure, the collimating plate 500 has a pair of horizontal plates 512 and 512′. The horizontal plates 512 and 512′ are rectangular plates and are constructed of an X-ray absorbing material such as lead for example. The horizontal plates 512 and 512′ lie on the same plane and their long sides are parallel to each other, while their short sides corresponding to each other lie on the same straight lines respectively. The horizontal plates 512 and 512′ are displaceable in their short side direction (x direction), whereby a distance “a” between their mutually opposed end faces can be adjusted. The horizontal plates 512 and 512′ are an example of a mode for carrying out the invention with respect to the first plate members defined herein. The horizontal plates 512 and 512′ are position-adjustable independently of each other. An example of a drive mechanism which permits such a positional adjustment is shown in FIG. 3. As shown in the same figure, the horizontal plates 512 and 512′ have arms 601 and 601′, respectively, which extend in y direction. The arms 601 and 601′ are engaged at end portions thereof with shafts 603 and 603′, respectively. The shafts 603 and 603′ are parallel shafts extending in x direction and spaced a predetermined interval from each other in z direction. The arm 601′ is bent to equalize the height of the horizontal plate 512′ with that of the horizontal plate 512 in z direction. The shafts 603 and 603′ are threaded throughout the overall length thereof. Engaging portions of the arms 601 and 601′ with the shafts 603 and 603′ are correspondingly threaded internally. Motors 605 and 605′ are mounted on one ends of the shafts 603 and 603′, respectively. The motor 605 is a reverse-rotatable motor. The motors 605 and 605′ are controlled each independently by a control means (not shown). A pair of follow-up plates 514 and 514′ are connected to the horizontal plates 512 and 512′, respectively. One ends of the follow-up plates 514 and 514′ are connected through hinges 516 and 516′ respectively to end faces of the horizontal plates 512 and 512′ on the side opposite to the mutually opposed end faces of the horizontal plates, while their opposite ends are engaged in guide grooves 520 and 520′ through pins 518 and 518′, respectively. The pins 518 and 518′ project in y direction from the opposite ends of the follow-up plates 514 and 514′. The guide grooves 520 and 520′ with those pins engaged therein are formed in a plate (not shown) which is opposed to the end faces in y direction of the follow-up plates 514 and 514′. The guide grooves 520 and 520′ are oblique to have a shape of inwardly inclined two lines in an xz plane. The inclination of the guide grooves 520 and 520′ is determined so as to run along the outer periphery of the quadrangular pyramid beam formed by the X-ray diaphragm 3. The pair of follow-up plates 514 and 514′ are also constructed of an X-ray absorbing material such as lead for example. The hinges 516 and 516′ are also formed of lead for example to block the passage of X-ray. The follow-up plates 514 and 514′ are one example of a mode for carrying out the invention with respect to the second plate member of the invention. A window plate 530 is disposed horizontally below the follow-up plates 514 and 514′. The window plate 530 has a square window. As to the size of the window, it takes a fixed value “b” in y direction, while in x direction it takes a fixed value larger than a maximum value of the spacing “a” between the horizontal plates 512 and 512′. With the collimating plate 500 of such a construction, as shown in FIG. 4, there is formed a quadrangular aperture having a size of a×b for the X-ray emitted from the X-ray tube 1. Since the positions of the horizontal plates 512 and 512′ can be changed, the size “a” in x direction of the aperture can be adjusted as desired. As shown in FIG. 5, the aperture can be set to a desired value a1 by adjusting the spacing between the horizontal plates 512 and 512′. Since the horizontal plates 512 and 512′ can be adjusted their positions each independently, it is possible to change the position of the aperture while maintaining the same opening, for example as shown in FIG. 6. By making end faces of the horizontal plates 512 and 512′ come closely into contact with each other, the aperture can be fully closed, as shown in FIG. 7. The full closing of the aperture can also be done in an offset state of the horizontal plates 512 and 512′, for example as shown in FIG. 8. The follow-up plates 514 and 514′ also move following the movement of the horizontal plates 512 and 512′ to cover the other portion than the aperture. In this case, the movement of lower end portions of the follow-up plates 514 and 514′ is restricted by the engagement of the pins 518 and 518′ with the guide grooves 520 and 520′, so that the movement of the follow-up plates 514 and 514′ is restricted so that their lower ends move along the guide grooves 520 and 520′. As a result, when the horizontal plates 512 and 512′ are moved to be close to each other, the lower ends of the follow-up plates 514 and 514′ rise along the guide grooves 520 and 520′, while when the horizontal plates 512 and 512′ are moved away from each other, the lower ends of the follow-up plates go down along the guide grooves 520 and 520′. When the horizontal plates 512 and 512′ are moved to a maximum limit away from each other, the aperture becomes fully open. This state is shown in FIG. 9. As shown in the same figure, the horizontal plates 512 and 512′ retreat to a maximum limit in the right and left direction to form a maximum aperture a0. At this time, the lower ends of the follow-up plates 514 and 514′ assume a lowest state in the guide grooves 520 and 520′. Thus, the lower ends of the follow-up plates 514 and 514′ move obliquely along the guide grooves 520 and 520′, so that even if the aperture is made fully open, the external shape of the collimating plate 500 does not become so large. Consequently, it is possible to reduce the size of the collimator 5 having the collimating plate 500 in the interior thereof. Besides, no special material is needed because the horizontal plates 512 and 512′ and the follow-up plates 514 and 514′ can each be fabricated using a lead plate or the like. Many widely different embodiments of the invention may be configured without departing from the spirit and the scope of the present invention. It should be understood that the present invention is not limited to the specific embodiments described in the specification, except as defined in the appended claims. |
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055679528 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS FIG. 3 shows a container with a non removable base in accordance with the present invention. FIG. 3a shows an enlargement of the base--tube interface. FIG. 3b shows a detail of the base--tube interface in the particular instance where the base is recessed within the base. Reference numerals 1, 2, 3, 4, 5, 6, 8, 10, 18, 19 have the same meanings as in FIGS. 1 and 2 of the prior art. Base (6) is located inside tube (1), of steel or other strong metal, and has a peripheral lateral wall (12) which is cylindrical with a circular cross section and is enclosed over its entire height by an identical cylindrical portion (13) machined in the internal wall (20) of tube (1). In the figure, the external surface (14) of base (6) does not extend beyond the plane containing the end face (10) of tube (1). Base (6) is then held in place by shrink fitting using the tube (1) itself. The cavity (22) of the container can be of any shape (for example with a polygonal cross section) such that its internal wall (20) requires a countersink (21) (FIG. 3a) to hold the base in place while other cavity (22) shapes (for example a circular cross section) would not require this countersink (see FIG. 3, 18a). The following can also be seen: the external weld seem (17) connects the peripheral edge of the external surface (14) of base (6) to the tube; PA1 the internal weld seam (18, 18a) which connects the peripheral edge of the internal surface (19) of base (6) to the tube; and at (18a), the internal wall (20) of the cavity has no countersink (as described above). Two cooperating shoulders (15) can also be seen, one machined in the inner wall of tube (1) and the other in the lateral (12) of base (6). It can be seen from FIG. 3b that the external surface (14) of base (6) is not flush with the end face (10) of tube (1), but is within it: this forms a circular disk which can, for example, be used to hold additional incompressible neutron shielding (23). The sealing weld (17) thus connects the external surface (14) of base (6) to the internal surface of tube (1). It should be noted that it could be advantageous to provide an allowance for assembly between base (6) and tube (1) in the vertical larger diameter portion (arrow 13) from the shoulder and it is preferable that this portion is of reduced height with respect to the smaller diameter portion (arrow 12). This is preferably sufficient to hold the base (6) in place even if it forms no part of the secure shrink fitting. It should also be noted that, as described above, the countersink (21) of the tube can be partially or completely absent over the internal surface of the tube and may or may not coexist with shoulder (15). It may also replace shoulder (15). |
claims | 1. A device configured to produce radioisotopes by irradiating a target fluid using a particle beam, the target fluid comprising a radioisotope precursor, the device comprising:an irradiation cell comprising:a conical cavity configured to contain the target fluid, the cavity having an opening at a base of the conical cavity, where the cavity base is surrounded by a front surface of the irradiation cell; anda metal foil connected to the front surface of the irradiation cell and closing the opening of the cavity, wherein the metal foil has a diameter less than or substantially equal to a diameter of the cavity base,wherein an outer surface of the conical cavity comprises furrows extending from an area close to an apex of the conical cavity toward a region close to the base of the cavity, so as to create pathways for the passage of non-cryogenic coolant to flow along the outer surface;a cooling device configured to circulate the non-cryogenic coolant and to cool the walls of the cavity; andan inclined surface, defining the bottom surface of the cavity, so as to evacuate the target fluid, which condenses in contact with the cavity walls, by gravity toward the metal foil;wherein the inclined surface intersects a plane formed by the metal foil at an acute angle (α) with the plane, so as to form, with the metal foil, a corner-shaped area that collects the evacuated target fluid, such that a height of the collected target fluid is maximal at the metal foil and decreases in a direction away from the metal foil. 2. The device according to claim 1, wherein the metal foil is positioned substantially perpendicular to an axis of the particle beam. 3. The device according to claim 1, wherein the radioisotopes are produced by irradiating a target fluid using a substantially horizontal particle beam. 4. The device according to claim 1, wherein a size of the acute angle (α) is between 30° and 89°. 5. The device according to claim 1, wherein the cooling device comprises:a coolant intake situated across from the part of the irradiation cell opposite the foil; anda diffuser creating a channel disposed to circulate the non-cryogenic coolant. 6. The device according to claim 1, wherein an apex of conical cavity is rounded. 7. The device according to claim 1, wherein the irradiation cell comprises:a first part comprising a front surface, which forms a bearing surface for the metal foil, and a rear surface; anda second, substantially conical part, which protrudes relative to the rear surface of the first part;wherein the cavity:passes through the first part to extend into the second part, andforms, in the front surface of the first part, an opening defined by an edge, such that the metal foil closes the opening at the edge when the metal foil bears on the front surface of the first part. 8. The device according to claim 7, wherein the first part further comprises a groove surrounding the second part on a side of the rear surface, the groove being configured to collect the non-cryogenic coolant flowing along an outer surface of the second part. 9. The device according to claim 1, wherein the irradiation cell is made from niobium. 10. An irradiation cell configured to produce radioisotopes by irradiating a target fluid using a particle beam, the target fluid comprising a radioisotope precursor, the irradiation cell comprising:a metal foil;a first part comprising a front surface and a rear surface, the front surface forming a bearing surface for the metal foil;a second, substantially conical part, which protrudes relative to the rear surface of the first part; anda substantially conical cavity, the cavity:having a bottom surface defined by an inclined plane;having an opening at a base of the conical cavity, where the cavity base is surrounded by a front surface of the irradiation cell;being configured to contain the target fluid;passing through the first part to extend into the second part; andrunning into the front surface of the first part at an acute angle (α) to form in the first part the opening defined by an edge,wherein an outer surface of the second part comprises furrows extending from an area close to an apex of the second part toward a region near a base of the second part, so as to create pathways between the furrows for the passage of a non-cryogenic coolant flowing along the outer surface of the second part, andwherein the metal foil is:connected to the front surface of the irradiation cell; andconfigured to close the opening at the edge when the metal foil bears on the front surface of the first part. 11. The irradiation cell according to claim 10, wherein the first part further comprises a groove, which, on a side of the rear surface of the first part, surrounds an outer surface of the second part, so as to reduce a thickness of the first part at the base of the second part, the groove being configured to collect the non-cryogenic coolant flowing along the outer surface of the second part. 12. The device according to claim 1, wherein the acute angle (α) has a size of between 45° and 85°. 13. The device according to claim 1, wherein the acute angle (α) has a size of between 60° and 85°. 14. The device according to claim 1, wherein the cavity comprises an inlet channel disposed proximal to the base of the cavity, the inlet channel being configured to introduce the target fluid into the cavity. 15. The device according to claim 1, wherein the inclined surface comprises an output channel disposed proximal to the base of the cavity, the output channel being configured to remove the collected target fluid. 16. The device according to claim 15, wherein the output channel is angled. 17. The device according to claim 1, wherein the cooling device comprises a diffuser forming an annular channel around the irradiation cell, the annular channel being configured to circulate the non-cryogenic coolant to cool walls of the cavity. |
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040010783 | description | Very generally, the nuclear reactor system incorporating the invention includes a pressure vessel 11 with at least one penetration 12 therein and a core region 13 enclosed by the pressure vessel and in which a plurality of control rod guide holes 14 are distributed over an area larger than the cross section of the penetration. Flexible control rods 16 are provided, one for each of the guide holes for insertion therein. A plurality of guide tubes 17 extend from the reactor vessel penetrations and fan out to respective ones of the guide holes for guiding the control rods from the penetration to the guide holes. Suitable means 18 are provided for moving the control rods through the guide tubes and into and out of the guide holes. A particular form of nuclear reactor system in which the invention is illustrated and described herein employs a reactive core comprised of a plurality of core blocks. The core blocks are stacked in columns and may be comprised of fissile or fertile material, neutron moderating material, neutron reflecting material, or combinations of some or all of these. A reactor core of this general type is shown and described in U.S. Pat. No. 3,359,175 assigned to the United States of America as represented by the Atomic Energy Commission. Although shown and described herein with a nuclear reactor system of the described type, the invention is applicable to other types of systems wherein a plurality of control rod guide holes are distributed over an area of the core larger than the cross section of the penetration which serves them. Referring more particularly to the drawings, one of the core regions 13 of which the reactor core is comprised is illustrated in connection with the apparatus of the invention. The reactor core may include a plurality of such core regions, each of which includes a group of seven fuel element columns of hexagonal cross section, each column being comprised of a plurality of core blocks. The core blocks are indicated at 21 and may be, as mentioned, fissile material, fertile material, reflector material, etc. The blocks may be held together with suitable dowel pins or interlocking configurations as described more particularly in the aforementioned U.S. patent. Each core region in the illustrated reactor system is comprised of seven columns of blocks, that is, a central column surrounded by six peripheral columns. The core blocks are provided with a plurality of aligned longitudinal passages therein indicated at 23 through which the reactor coolant passes. Each of the core blocks is provided with a central hole 14 therein which is formed with a shoulder 25. A suitable grappling tool, not illustrated, may be inserted in the holes 14 and expanded to abut the shoulder 25 for lifting the blocks out of the reactor core during refueling procedures. The holes 14 in the peripheral columns are also used for accommodating the control rod, as will be explained. Each of the core regions 13 is provided with a plenum 27 at the top thereof. The plenum 27 is formed by the outer walls 29 of a series of partially hexagonal elements 30 attached to the upper surface of the core region 13. The upper parts of the elements 30 contain material which forms an annular radiation shield 31 above the plenum 27. The plenum opens inwardly on the region 33 defined by it and the annular shield. The inner walls of the top parts of the elements 30 are arcuate and form a circular wall 32 bordering on the region 33. A tube 35 is provided at the top of each of the holes 14 in the peripheral columns to provide an extension of the hole through the plenum to the top plate 37 of the housing. In the reactor system, the space above the reactor core and below the pressure vessel upper wall is pressurized by the flow of coolant gas. Regulation of the flow of pressurized gas from this region into the plenum 27 is controlled by a suitable valve positioned in and above the region 33 formed within the plenum 27 and shield 31 and above the central column of core blocks in each core region. Although the valve may be of any suitable configuration, the illustrated embodiment includes a plate 41 extending horizontally above the region 33 and spaced therefrom. The plate 41 is supported in a block 43 which is attached to a tube 45. The tube 45 rests on a bearing pad 47 which rests on the top surface of the central column of the core region aligned with the central hole 14 therein. The tube 45 extends upwardly through the penetration 12. A sleeve 46 is attached to a segmented frustoconical section 49 and extends upwardly coaxial with the tube 45. The frustoconical section 49 serves as a guide to facilitate removal of the apparatus through the penetration 12 in the reactor vessel 11. With the valve element in its full downward or open position as shown, coolant is able to flow into the region 33 and the plenum 27 through the openings or windows 53 provided in the valve element 48. The lower edges 55 of the windows 53 are curved for fine flow adjustment. Guide plates 57 extend inwardly from the valve element 48 to assist in assembly. The vertical position of the valve element 48 with respect to the core column is adjusted by moving the sleeve 46 to vary the area of the windows 53 above the upper surface 37 of the elements 30 and thus regulate the flow of coolant into the region 33 and the plenum 27. The sleeve 46 connects the valve element 48 to a mechanism (not shown) exteriorly of the center cavity of the pressure vessel. The pressure vessel penetration 12 is aligned with the central column in the core region 13. The penetration is provided with a penetration liner 61 and the control rod drive mechanism is mounted in a cylindrical housing 63 supported within the penetration 12. A lower plate 65 extends across the housing 63. The entire assembly of the valve 48, control rods 16, guides 17, housing 63, and connecting structure may be withdrawn in its entirety from the penetration and replaced with a suitable fuel element handling system during refueling operations. Control rod drive guide tubes 67 are supported within the housing 63 and extend axially within the housing, each of the tubes 67 corresponding to one of the peripheral columns of the core region 13. The control rods which move through the holes 14 in the peripheral columns may be withdrawn into the guide tubes 67 when it is desired to maximize the neutron flux in the reactor core. Preferably, at least some of the control rods in each column are individually controllable in their positions. Because the holes 14 of the reactor core are arrayed about a circle having a diameter substantially larger than the diameter of the reactor vessel penetration, means are provided for enabling the control rods 16 to pass freely from the holes 14 into the guide tubes 67. To this end, the lower ends of the guide tubes 67 protrude into the space above the reactor core and are provided with a flexible bellows 69 on the lower projecting ends thereof. A more detailed view of the bellows may be seen in FIG. 4. An alternative design may utilize a hinge coupling if desired. Guide tubes 17 extend from each of the bellows 69 to the upper ends of the respective tubes 35 forming extensions of the holes 14. The tubes 17 fan outwardly to join the more widely spaced holes with the more closely groupled guide tubes 67. The lower ends of the guide tubes 17 are provided with bushings 73 which seal in suitable bushings 75 attached to the upper ends of the tubes 35. In order to pass through the tubes 17, the control rods 16 are made flexible. In the illustrated embodiment, this is accomplished by manufacturing the control rods of a plurality of segments 77. Each of the segments in joined together, as shown in FIG. 3, by means of a rod extension 79 to the end of which is attached a semi-hemispherical ball 80. A ball housing 83 extends from the adjacent segment and captures the hemispherical ball 80 to provide the desired flexibility. In the case of a gas-cooled reactor, it is typically desirable to coat all rubbing and contacting surfaces with chromium carbide or the equivalent to prevent galling and self-welding due to the dryness of the coolant gas. As previously mentioned, during refueling operations the control rod drive assembly and associated elements are removed from the penetration 12. In order to allow the apparatus to clear the penetration, the control rods are first withdrawn into the tubes 67, 17 and 45 within the housing 63 but clear of the bushings 75 and 47. The tubes 17 are then retracted upward and inward to fit within the penetration in the reactor vessel. This is accomplished by means of a longitudinal sleeve 81 coaxial with the sleeve 86 and the sleeve 46 which controls the valve for the reactor coolant. The sleeve 81 extends through a bearing sleeve 87 mounted on a spider-like support 85 extending from the lower ends of the tubes 67. A collar 88 fixed to the tube 86 near its lower end is attached to each of the tubes 17 by means of a link 89 which extends from a mounting bracket 91 on the tubes and is pivotally attached at each end. To remove the guide tubes 17 for refueling, the sleeve 81 and the tube 86 are drawn upwardly together until the bushings 73 clear the plate 41. The tube 86 is then moved upwardly relative to the sleeve 81 and actuates the linkage 88, 89 and 91. The linkage operates to move the tubes 17 through slots in the guide 49 to the attitude shown in phantom, thus allowing the tubes to be withdrawn upwardly through the reactor pressure vessel penetration. A suitable mechanism, not shown, is provided outside the reactor vessel for accomplishing the foregoing movements. The reverse process is performed when the control rods arrangement is replaced after refueling. In addition to the structural advantages accruing from the invention, a number of advantages occur also in the reactor physics. The use of a large number of small lowworth rods in a reactor core permits greater flexibility in rod programming schemes over the use of fewer and larger diameter rods. By selection of control rod insertion patterns, radial power distribution and shape can be selected as desired. Moreover, a more uniformly distributed rod pattern minimizes the perturbation of the power distribution which movement of the rods induces. Variations in power density as a result of aging of the various fuel elements at different rates is more readily controlled. Moreover, the axial power distribution can also be affected by judicious choice of the individual rod insertion pattern. Higher power can be tolerated for the same fuel temperature or a lower fuel temperature can be used to ensure a lower fuel failure rate with subsequent reduced circulating fission product activity as a result of a more even radial and axial distribution of neutron flux and power generation rate. A more uniform temperature distribution also reduces radiation induced stress and distortion. When graphite is exposed to fast neutron bombardment, it experiences a densification which results in dimensional changes. The process is a function of the fast neutron dose and the temperature of the graphite during the exposure. When a temperature gradient exists in the graphite, differential strain produces internal stress and distortion. With reduced temperature and flux gradients, less differential strain and thereby less stress and overall distortion occurs. A control rod may also be used in the central column. The central column rod may be used as a minimal reactivity "shim" rod so as to more evenly make changes in the power level of the reactor. The advantages to the axial power peaking are the same as for the other six control rods in the peripheral column. In addition, a hole may be provided in the central column to provide reserve shut-down capacity, or rather than a single hole, two reserve shut-down holes symmetrically located in the central column may be used. Tubes for conveying emergency shut-down poison are shown in FIG. 1 at 92. Because of the large number of control rods of the present invention, sufficient shut-down capacity may be available so as to eliminate the need for the tubes 92. Under such circumstances, more advantages are available. The fuel blocks can be all of the same design, reducing inventory problems. Also, the absence of the holes for the tubes 92 provides more volume in the core for fuel and power generating capacity. Other advantages result from the invention in terms of the greater flexibility of operation due to the large number of control rods and the ability to more closely regulate flux distribution and burn-up rates with respect to the various columns. It may therefore be seen that the invention provides an improved nuclear reactor system and control apparatus therefor. Control rods may be inserted into holes in the reactor core which are spaced about an area which is larger than the cross section of the reactor pressure vessel penetration through which the control rods are operated. As a result, the integrity of the pressure vessel is maximized while at the same time ensuring a large distribution of control throughout the reactor core. Various modifications of the invention in addition to those shown and described herein will become apparent to those skilled in the art from the foregoing description and accompanying drawings. Such modifications are intended to fall within the scope of the appended claims. |
abstract | A nuclear reactor is positioned on a barge which is floating in a water tank. A plurality of counter weight assemblies interconnect the barge with the tank to create a lifting force to the barge and to maintain the barge in a level position. Structure is also included for limiting horizontal movement of the counter weight of the counter weight assemblies. |
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