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description | This application is a continuation of U.S. patent application Ser. No. 12/264,854, filed Nov. 4, 2008, which is a continuation of U.S. patent application Ser. No. 11/695,532, filed Apr. 2, 2007, issued as U.S. Pat. No. 7,446,328 on Nov. 4, 2008, which is a continuation of U.S. patent application Ser. No. 10/917,023, filed Aug. 12, 2004, issued as U.S. Pat. No. 7,199,382 on Apr. 3, 2007, which claims the benefit of U.S. Provisional Application No. 60/494,699, filed Aug. 12, 2003, and U.S. Provisional Application No. 60/579,095, filed Jun. 10, 2004, both entitled “Precision Patient Alignment and Beam Therapy System.” This invention was made with United States Government support under the DAMD17-99-1-9477 and DAMD17-02-1-0205 grants awarded by the Department of Defense. The Government has certain rights in the invention. 1. Field of the Invention The invention relates to the field of radiation therapy systems. One embodiment includes an alignment system with an external measurement system and local feedback to improve accuracy of patient registration and positioning and to compensate for misalignment caused by factors such as mechanical movement tolerances and non-strictly rigid structures. 2. Description of the Related Art Radiation therapy systems are known and used to provide treatment to patients suffering a wide variety of conditions. Radiation therapy is typically used to kill or inhibit the growth of undesired tissue, such as cancerous tissue. A determined quantity of high-energy electromagnetic radiation and/or high-energy particles is directed into the undesired tissue with the goal of damaging the undesired tissue while reducing unintentional damage to desired or healthy tissue through which the radiation passes on its path to the undesired tissue. Proton therapy has emerged as a particularly efficacious treatment for a variety of conditions. In proton therapy, positively charged proton subatomic particles are accelerated, collimated into a tightly focused beam, and directed towards a designated target region within the patient. Protons exhibit less lateral dispersion upon impact with patient tissue than electromagnetic radiation or low mass electron charged particles and can thus be more precisely aimed and delivered along a beam axis. Also, upon impact with patient tissue, the accelerated protons pass through the proximal tissue with relatively low energy transfer and then exhibit a characteristic Bragg peak wherein a significant portion of the kinetic energy of the accelerated mass is deposited within a relatively narrow penetration depth range within the patient. This offers the significant advantage of reducing delivery of energy from the accelerated proton particles to healthy tissue interposed between the target region and the delivery nozzle of a proton therapy machine as well as to “downrange” tissue lying beyond the designated target region. Depending on the indications for a particular patient and their condition, delivery of the therapeutic proton beam may preferably take place from a plurality of directions in multiple treatment fractions to achieve a total dose delivered to the target region while reducing collateral exposure of interposed desired/healthy tissue. Thus, a radiation therapy system, such as a proton beam therapy system, typically has provision for positioning and aligning a patient with respect to a proton beam in multiple orientations. In order to determine a preferred aiming point for the proton beam within the patient, the typical procedure has been to perform a computed tomography (CT) scan in an initial planning or prescription stage from which multiple digitally reconstructed radiographs (DRRs) can be determined. The DRRs synthetically represent the three dimensional data representative of the internal physiological structure of the patient obtained from the CT scan in two dimensional views considered from multiple orientations and thus can function as a target image of the tissue to be irradiated. A desired target isocenter corresponding to the tissue to which therapy is to be provided is designated. The spatial location of the target isocenter can be referenced with respect to physiological structure of the patient (monuments) as indicated in the target image. Upon subsequent setup for delivery of the radiation therapy, a radiographic image is taken of the patient, such as a known x-ray image, and this radiographic image is compared or registered with the target image with respect to the designated target isocenter. The patient's position is adjusted to, as closely as possible or within a given tolerance, align the target isocenter in a desired pose with respect to the radiation beam as indicated by the physician's prescription. The desired pose is frequently chosen as that of the initial planning or prescription scan. In order to reduce misalignment of the radiation beam with respect to the desired target isocenter to achieve the desired therapeutic benefit and reduce undesired irradiation of other tissue, it will be appreciated that accuracy of placement of the patient with respect to the beam nozzle is important to achieve these goals. In particular, the target isocenter is to be positioned translationally to coincide with the delivered beam axis as well as in the correct angular position to place the patient in the desired pose in a rotational aspect. In particular, as the spatial location of the Bragg peak is dependent both upon the energy of the delivered proton beam as well as the depth and constitution of tissue through which the beam passes, it will be appreciated that a rotation of the patient about the target isocenter even though translationally aligned can present a varying depth and constituency of tissue between the initial impact point and the target isocenter located within the patient's body, thus varying the penetration depth. A further difficulty with registration and positioning is that a radiation therapy regimen typically is implemented via a plurality of separate treatment sessions administered over a period of time, such as daily treatments administered over a several week period. Thus, the alignment of the patient and the target isocenter as well as positioning of the patient in the desired pose with respect to the beam is typically repeatedly determined and executed multiple times over a period of days or weeks. There are several difficulties with accurately performing this patient positioning with respect to the radiation treatment apparatus. As previously mentioned, patient registration is performed by obtaining radiographic images of the patient at a current treatment session at the radiation therapy delivery site and comparing this obtained image with the previously obtained DRR or target image which is used to indicate the particular treatment prescription for the patient. As the patient will have removed and repositioned themselves within the radiation therapy apparatus, the exact position and pose of a patient will not be exactly repeated from treatment session to treatment session nor to the exact position and pose with which the target image was generated, e.g., the orientation from which the original CT scan generated the DRRs. Thus, each treatment session/fraction typically involves precisely matching a subsequently obtained radiographic image with an appropriate corresponding DRR to facilitate the determination of a corrective translational and/or rotational vector to position the patient in the desired location and pose. In addition to the measurement and computational difficulties presented by such an operation, is the desire for speed in execution as well as accuracy. In particular, a radiation therapy apparatus is an expensive piece of medical equipment to construct and maintain both because of the materials and equipment needed in construction and the indication for relatively highly trained personnel to operate and maintain the apparatus. In addition, radiation therapy, such as proton therapy, is increasingly being found an effective treatment for a variety of patient conditions and thus it is desirable to increase patient throughput both to expand the availability of this beneficial treatment to more patients in need of the same as well as reducing the end costs to the patients or insurance companies paying for the treatment and increase the profitability for the therapy delivery providers. As the actual delivery of the radiation dose, once the patient is properly positioned, is a relatively quick process, any additional latency in patient ingress and egress from the therapy apparatus, imaging, and patient positioning and registration detracts from the overall patient throughput and thus the availability, costs, and profitability of the system. A further difficulty with accurately positioning the patient and the corresponding target isocenter in the desired position and pose with respect to the beam nozzle are the multiple and additive uncertainties in the exact position and relative angle of the various components of a radiation therapy system. For example, the beam nozzle can be fitted to a relatively rigid gantry structure to allow the beam nozzle to revolve about a gantry center to facilitate presentation of the radiation beam from a variety of angles with respect to the patient without requiring uncomfortable or inconvenient positioning of the patient themselves. However, as the gantry structure is relatively large (on the order of several meters), massive, and made out of non-strictly rigid materials, there is inevitably some degree of structural flex/distortion and non-repeatable mechanical tolerance as the nozzle revolves about the gantry. Further, the nozzle may be configured as an elongate distributed mass that is also not strictly rigid such that the distal emissions end of the nozzle can flex to some degree, for example as the nozzle moves from an overhead vertical position to a horizontal, sideways presentation of the beam. Accurate identification of the precise nozzle position can also be complicated by a cork screwing with the gantry. Similarly, the patient may be placed on a supportive pod or table and it may be connected to a patient positioning apparatus, both of which are subject to some degree of mechanical flex under gravity load, as well as mechanical tolerances at moving joints that are not necessarily consistent throughout the range of possible patient postures. While it is possible to estimate and measure certain of these variations, as they are typically variable and non-repeatable, it remains a significant challenge to repeatedly position a patient consistently over multiple treatment sessions in both location and pose to tight accuracy limits, such as to millimeter or less accuracy on a predictive basis. Thus, the known way to address gantry and patient table misalignment is to re-register the patient before treatment. This is undesirable as the patient is exposed to additional x-ray radiation for the imaging and overall patient throughput is reduced by the added latency of the re-registration. From the foregoing it will be understood that there is a need for increasing the accuracy and speed of the patient registration process. There is also a need for reducing iteratively imaging and reorienting the patient to achieve a desired pose. There is also a need for a system that accounts for variable and unpredictable position errors to increase the accuracy of patient registration and alignment with a radiation therapy delivery system. Embodiments of the invention provide a patient alignment system that externally measures and provides corrective feedback for variations or deviations from nominal position and orientation between the patient and a delivered therapeutic radiation beam. The alignment system can readily accommodate variable and unpredictable mechanical tolerances and structural flex of both fixed and movable components of the radiation therapy system. The patient alignment system reduces the need for imaging the patient between treatment fractions and decreases the latency of the registration process, thus increasing patient throughput. Other embodiments comprise a radiation therapy delivery system comprising a gantry, a patient fixation device configured to secure a patient with respect to the patient fixation device, a patient positioner interconnected to the patient fixation device so as to position the patient fixation device along translational and rotational axes within the gantry, a radiation therapy nozzle interconnected to the gantry and selectively delivering radiation therapy along a beam axis, a plurality of external measurement devices which obtain position measurements of at least the patient fixation device and the nozzle, and a controller which receives the position measurements of at least the patient fixation device and the nozzle and provides control signals to the patient positioner to position the patient in a desired orientation with respect to the beam axis. Another embodiment comprises a patient positioning system for a radiation therapy system having a plurality of components that are subject to movement, the positioning system comprising a plurality of external measurement devices arranged to obtain position measurements of the plurality of components so as to provide location information, a movable patient support configured to support a patient substantially fixed in position with respect to the patient support and controllably position the patient in multiple translational and rotational axes, and a controller receiving information from the plurality of external measurement devices and providing movement commands to the movable patient support to align the patient in a desired pose such that the positioning system compensates for movement of the plurality of components. Further embodiments include a method of registering and positioning a patient for delivery of therapy with a system having a plurality of components subject to movement, the method comprising the steps of positioning a patient in an initial treatment pose with a controllable patient positioner, externally measuring the location of selected points of the plurality of components, determining a difference vector between the observed initial patient pose and a desired patient pose, and providing movement commands to the patient positioner to bring the patient to the desired patient pose. Yet another embodiment comprises a positioning system for use with a radiation treatment facility wherein the radiation treatment facility has a plurality of components that includes a source of particles and a nozzle from which the particles are emitted, wherein the nozzle is movable with respect to the patient to facilitate delivery of the particles to a selected region of the patient via a plurality of different paths, the positioning system comprising a patient positioner that receives the patient wherein the patient positioner is movable so as to orient the patient with respect to the nozzle to facilitate delivery of the particles in the selected region of the patient, a monitoring system that images at least one component of the radiation treatment facility in proximity to the patient positioner, wherein the monitoring system develops a treatment image indicative of the orientation of the at least one component with respect to the patient prior to treatment, and a control system that controls delivery of particles to the patient wherein the control system receives signals indicative of the treatment to be performed, the signals including a desired orientation of the at least one component when the particles are to be delivered to the patient, wherein the control system further receives the treatment image and the control system evaluates the treatment image to determine an actual orientation of the at least one component prior to treatment and wherein the control system compares the actual orientation of the at least one component prior to treatment to the desired orientation of the at least one component and, if the actual orientation does not meet a pre-determined criteria for correspondence with the desired orientation, the control system sends signals to the patient positioner to move the patient positioner such that the actual orientation more closely corresponds to the desired orientation during delivery of the particles. These and other objects and advantages of the invention will become more apparent from the following description taken in conjunction with the accompanying drawings. Reference will now be made to the drawings wherein like reference designators refer to like parts throughout. FIGS. 1A and 1B illustrate schematically first and second orientations of one embodiment of a radiation therapy system 100, such as based on the proton therapy system currently in use at Loma Linda University Medical Center in Loma Linda, Calif. and as described in U.S. Pat. No. 4,870,287 of Sep. 26, 1989 which is incorporated herein in its entirety by reference. The radiation therapy system 100 is designed to deliver therapeutic radiation doses to a target region within a patient for treatment of malignancies or other conditions from one or more angles or orientations with respect to the patient. The system 100 includes a gantry 102 which includes a generally hemispherical or frustoconical support frame for attachment and support of other components of the radiation therapy system 100. Additional details on the structure and operation of embodiments of the gantry 102 may be found in U.S. Pat. No. 4,917,344 and U.S. Pat. No. 5,039,057, both of which are incorporated herein in their entirety by reference. The system 100 also comprises a nozzle 104 which is attached and supported by the gantry 102 such that the gantry 102 and nozzle 104 may revolve relatively precisely about a gantry isocenter 120, but subject to corkscrew, sag, and other distortions from nominal. The system 100 also comprises a radiation source 106 delivering a radiation beam along a radiation beam axis 140, such as a beam of accelerated protons. The radiation beam passes through and is shaped by an aperture 110 to define a therapeutic beam delivered along a delivery axis 142. The aperture 110 is positioned on the distal end of the nozzle 104 and the aperture 110 may preferably be specifically configured for a patient's particular prescription of therapeutic radiation therapy. In certain applications, multiple apertures 110 are provided for different treatment fractions. The system 100 also comprises one or more imagers 112 which, in this embodiment, are retractable with respect to the gantry 102 between an extended position as illustrated in FIG. 2A and a retracted position as illustrated in FIG. 2B. The imager 112 in one implementation comprises a commercially available solid-state amorphous silicon x-ray imager which can develop image information such as from incident x-ray radiation that has passed through a patient's body. The retractable aspect of the imager 112 provides the advantage of withdrawing the imager screen from the delivery axis 142 of the radiation source 106 when the imager 112 is not needed thereby providing additional clearance within the gantry 102 enclosure as well as placing the imager 112 out of the path of potentially harmful emissions from the radiation source 106 thereby reducing the need for shielding to be provided to the imager 112. The system 100 also comprises corresponding one or more x-ray sources 130 which selectively emit appropriate x-ray radiation along one or more x-ray source axes 144 so as to pass through interposed patient tissue to generate a radiographic image of the interposed materials via the imager 112. The particular energy, dose, duration, and other exposure parameters preferably employed by the x-ray source(s) 130 for imaging and the radiation source 106 for therapy will vary in different applications and will be readily understood and determined by one of ordinary skill in the art. In this embodiment, at least one of the x-ray sources 130 is positionable such that the x-ray source axis 144 can be positioned so as to be nominally coincident with the delivery axis 142. This embodiment provides the advantage of developing a patient image for registration from a perspective which is nominally identical to a treatment perspective. This embodiment also includes the aspect that a first imager 112 and x-ray source 130 pair and a second imager 112 and x-ray source 130 pair are arranged substantially orthogonal to each other. This embodiment provides the advantage of being able to obtain patient images in two orthogonal perspectives to increase registration accuracy as will be described in greater detail below. The imaging system can be similar to the systems described in U.S. Pat. Nos. 5,825,845 and 5,117,829 which are hereby incorporated by reference. The system 100 also comprises a patient positioner 114 (FIG. 3) and a patient pod 116 which is attached to a distal or working end of the patient positioner 114. The patient positioner 114 is adapted to, upon receipt of appropriate movement commands, position the patient pod 116 in multiple translational and rotational axes and preferably is capable of positioning the patient pod 116 in three orthogonal translational axes as well as three orthogonal rotational axes so as to provide a full six degree freedom of motion to placement of the patient pod 116. The patient pod 116 is configured to hold a patient securely in place in the patient pod 116 so to as substantially inhibit any relative movement of the patient with respect to the patient pod 116. In various embodiments, the patient pod 116 comprises expandable foam, bite blocks, and/or fitted facemasks as immobilizing devices and/or materials. The patient pod 116 is also preferably configured to reduce difficulties encountered when a treatment fraction indicates delivery at an edge or transition region of the patient pod 116. Additional details of preferred embodiments of the patient positioner 114 and patient pod 116 can be found in the commonly assigned application (Ser. No. 10/917,022, filed Aug. 12, 2004) entitled “Modular Patient Support System” filed concurrently herewith and which is incorporated herein in its entirety by reference. As previously mentioned, in certain applications of the system 100, accurate relative positioning and orientation of the therapeutic beam delivery axis 142 provided by the radiation source 106 with target tissue within the patient as supported by the patient pod 116 and patient positioner 114 is an important goal of the system 100, such as when comprising a proton beam therapy system. However, as previously mentioned, the various components of the system 100, such as the gantry 102, the nozzle 104, radiation source 106, the imager(s) 112, the patient positioner 114, the patient pod 116, and x-ray source(s) 130 are subject to certain amounts of structural flex and movement tolerances from a nominal position and orientation which can affect accurate delivery of the beam to that patient. FIGS. 1A and 1B illustrate different arrangements of certain components of the system 100 and indicate by the broken arrows both translational and rotational deviations from nominal that can occur in the system 100. For example, in the embodiment shown in FIG. 1A, the nozzle 104 and first imager 112 extend substantially horizontally and are subject to bending due to gravity, particularly at their respective distal ends. The second imager 112 is arranged substantially vertically and is not subject to the horizontal bending of the first imager 112. FIG. 1B illustrates the system 100 in a different arrangement rotated approximately 45° counterclockwise from the orientation of FIG. 1A. In this orientation, both of the imagers 112 as well as the nozzle 104 are subject to bending under gravity, but to a different degree than in the orientation illustrated in FIG. 1A. The movement of the gantry 102 between different orientations, such as is illustrated in FIGS. 1A and 1B also subjects components of the system 100 to mechanical tolerances at the moving surfaces. As these deviations from nominal are at least partially unpredictable, non-repeatable, and additive, correcting for the deviations on a predictive basis is extremely challenging and limits overall alignment accuracy. It will be appreciated that these deviations from the nominal orientation of the system are simply exemplary and that any of a number of sources of error can be addressed by the system disclosed herein without departing from the spirit of the present invention. FIGS. 4A-4E illustrate in greater detail embodiments of potential uncertainties or errors which can present themselves upon procedures for alignment of, for example, the nozzle 104 and the target tissue of the patient at an isocenter 120. FIGS. 4A-4E illustrate these sources of uncertainty or error with reference to certain distances and positions. It will be appreciated that the sources of error described are simply illustrative of the types of errors addressed by the system 100 of the illustrated embodiments and that the system 100 described is capable of addressing additional errors. In this embodiment, a distance SAD is defined as a source to axis distance from the radiation source 106 to the rotation axis of the gantry, which ideally passes through the isocenter 120. For purposes of explanation and appreciation of relative scale and distances, in this embodiment, SAD is approximately equal to 2.3 meters. FIG. 4A illustrates that one of the potential sources of error is a source error where the true location of the radiation source 106 is subject to offset from a presumed or nominal location. In this embodiment, the therapeutic radiation beam as provided by the radiation source 106 passes through two transmission ion chambers (TIC) which serve to center the beam. These are indicated as TIC 1 and TIC 3 and these are also affixed to the nozzle 104. The source error can arise from numerous sources including movement of the beam as observed on TIC 1 and/or TIC 3, error in the true gantry 102 rotational angle, and error due to “egging” or distortion from round of the gantry 102 as it rotates. FIG. 4A illustrates source error comprising an offset of the true position of the radiation source 106 from a presumed or nominal location and the propagation of the radiation beam across the SAD distance through the aperture 110 providing a corresponding error at isocenter 120. FIG. 4B illustrates possible error caused by TIC location error, where TIC 1, the radiation source 106, and TIC 3 are offset from an ideal beam axis passing through the nominal gantry isocenter 120. As the errors illustrated by FIGS. 4A and 4B are assumed random and uncorrelated, they can be combined in quadrature and projected through an assumed nominal center of the aperture 110 to establish a total error contribution due to radiation source 106 error projected to the isocenter 120. In this embodiment, before corrective measures are taken (as described in greater detail below), the radiation source error can range from approximately ±0.6 mm to ±0.4 mm. FIG. 4C illustrates error or uncertainty due to position of the aperture 110. The location of the radiation source 106 is assumed nominal; however, error or uncertainty is introduced both by tolerance stack-up, skew, and flex of the nozzle 104 as well as manufacturing tolerances of the aperture 110 itself. Again, as projected from the radiation source 106 across the distance SAD to the nominal isocenter 120, a beam delivery aiming point (BDAP) error is possible between a presumed nominal BDAP and an actual BDAP. In this embodiment, this BDAP error arising from error in the aperture 110 location ranges from approximately ±1.1 mm to ±1.5 mm. The system 100 is also subject to error due to positioning of the imager(s) 112 as well as the x-ray source(s) 130 as illustrated in FIGS. 4D and 4E. FIG. 4D illustrates the error due to uncertainty in the imager(s) 112 position with the position of the corresponding x-ray source(s) 130 assumed nominal. As the emissions from the x-ray source 130 pass through the patient assumed located substantially at isocenter 120 and onward to the imager 112, this distance may be different than the SAD distance and in this embodiment is approximately equal to 2.2 meters. Error or uncertainty in the true position of an imager 112 can arise from lateral shifts in the true position of the imager 112, errors due to axial shifting of the imager 112 with respect to the corresponding x-ray source 130, as well as errors in registration of images obtained by imager 112 to the DRRs. In this embodiment, before correction, the errors due to each imager 112 are approximately ±0.7 mm. Similarly, FIG. 4E illustrates errors due to uncertainty in positioning of the x-ray source(s) 130 with the position of the corresponding imager(s) 112 assumed nominal. Possible sources of error due to the x-ray source 130 include errors due to initial alignment of the x-ray source 130, errors arising from movement of the x-ray source 130 into and out of the beam line, and errors due to interpretation of sags and relative distances of TIC 1 and TIC 3. These errors are also assumed random and uncorrelated or independent and are thus added in quadrature resulting, in this embodiment, in error due to each x-ray source 130 of approximately ±0.7 mm. As these errors are random and independent and uncorrelated and thus potentially additive, in this embodiment the system 100 also comprises a plurality of external measurement devices 124 to evaluate and facilitate compensating for these errors. In one embodiment, the system 100 also comprises monuments, such as markers 122, cooperating with the external measurement devices 124 as shown in FIGS. 2A, 2B, 6 and 7. The external measurement devices 124 each obtain measurement information about the three-dimensional position in space of one or more components of the system 100 as indicated by the monuments as well as one or more fixed landmarks 132 also referred to herein as the “world” 132. In this embodiment, the external measurement devices 124 comprise commercially available cameras, such as CMOS digital cameras with megapixel resolution and frame rates of 200-1000 Hz, which independently obtain optical images of objects within a field of view 126, which in this embodiment is approximately 85° horizontally and 70° vertically. The external measurement devices 124 comprising digital cameras are commercially available, for example as components of the Vicon Tracker system from Vicon Motion Systems Inc. of Lake Forrest, Calif. However, in other embodiments, the external measurement devices 124 can comprise laser measurement devices and/or radio location devices in addition to or as an alternative to the optical cameras of this embodiment. In this embodiment, the markers 122 comprise spherical, highly reflective landmarks which are fixed to various components of the system 100. In this embodiment, at least three markers 122 are fixed to each component of the system 100 of interest and are preferably placed asymmetrically, e.g. not equidistant from a centerline nor evenly on corners, about the object. The external measurement devices 124 are arranged such that at least two external measurement devices 124 have a given component of the system 100 and the corresponding markers 122 in their field of view and in one embodiment a total of ten external measurement devices 124 are provided. This aspect provides the ability to provide binocular vision to the system 100 to enable the system 100 to more accurately determine the location and orientation of components of the system 100. The markers 122 are provided to facilitate recognition and precise determination of the position and orientation of the objects to which the markers 122 are affixed, however in other embodiments, the system 100 employs the external measurement devices 124 to obtain position information based on monuments comprising characteristic outer contours of objects, such as edges or corners, comprising the system 100 without use of the external markers 122. FIG. 5 illustrates one embodiment of determining the spatial position and angular orientation of a component of the system 100. As the component(s) of interest can be the gantry 102, nozzle 104, aperture 110, imager 112, world 132 or other components, reference will be made to a generic “object”. It will be appreciated that the process described for the object can proceed in parallel or in a series manner for multiple objects. Following a start state, in state 150 the system 100 calibrates the multiple external measurement devices 124 with respect to each other and the world 132. In the calibration state, the system 100 determines the spatial position and angular orientation of each external measurement device 124. The system 100 also determines the location of the world 132 which can be defined by a dedicated L-frame and can define a spatial origin or frame-of-reference of the system 100. The world 132 can, of course, comprise any component or structure that is substantially fixed within the field of view of the external measurement devices 124. Hence, structures that are not likely to move or deflect as a result of the system 100 can comprise the world 132 or point of reference for the external measurement devices 124. A wand, which can include one or more markers 122 is moved within the fields of view 126 of the external measurement devices 124. As the external measurement devices 124 are arranged such that multiple external measurement devices 124 (in this embodiment at least two) have an object in the active area of the system 100 in their field of view 126 at any given time, the system 100 correlates the independently provided location and orientation information from each external measurement device 124 and determines corrective factors such that the multiple external measurement devices 124 provide independent location and orientation information that is in agreement following calibration. The particular mathematical steps to calibrate the external measurement devices 124 are dependent on their number, relative spacing, geometrical orientations to each other and the world 132, as well as the coordinate system used and can vary among particular applications, however will be understood by one of ordinary skill in the art. It will also be appreciated that in certain applications, the calibration state 150 would need to be repeated if one or more of the external measurement devices 124 or world 132 is moved following calibration. Following the calibration state 150, in state 152 multiple external measurement devices 124 obtain an image of the object(s) of interest. From the images obtained in state 152, the system 100 determines a corresponding direction vector 155 to the object from each corresponding external measurement device 124 which images the object in state 154. This is illustrated in FIG. 6 as vectors 155a-d corresponding to the external measurement devices 124a-d which have the object in their respective fields of view 126. Then, in state 156, the system 100 calculates the point in space where the vectors 155 (FIG. 6) determined in state 154 intersect. State 156 thus returns a three-dimensional location in space, with reference to the world 132, for the object corresponding to multiple vectors intersecting at the location. As the object has been provided with three or more movements or markers 122, the system 100 can also determine the three-dimensional angular orientation of the object by evaluating the relative locations of the individual markers 122 associated with the object. In this implementation, the external measurement devices 124 comprise cameras, however, any of a number of different devices can be used to image, e.g., determine the location, of the monuments without departing from the spirit of the present invention. In particular, devices that emit or receive electromagnetic or audio energy including visible and non-visible wavelength energy and ultra-sound can be used to image or determine the location of the monuments. The location and orientation information determined for the object is provided in state 160 for use in the system 100 as described in greater detail below. In one embodiment, the calibration state 150 can be performed within approximately one minute and allows the system 100 to determine the object's location in states 152, 154, 156, and 160 to within 0.1 mm and orientation to within 0.15° with a latency of no more than 10 ms. As previously mentioned, in other embodiments, the external measurement devices 124 can comprise laser measurement devices, radio-location devices or other devices that can determine direction to or distance from the external measurement devices 124 in addition to or as an alternative to the external measurement devices 124 described above. Thus, in certain embodiments a single external measurement device 124 can determine both range and direction to the object to determine the object location and orientation. In other embodiments, the external measurement devices 124 provide only distance information to the object and the object's location in space is determined by determining the intersection of multiple virtual spheres centered on the corresponding external measurement devices 124. In certain embodiments, the system 100 also comprises one or more local position feedback devices or resolvers 134 (See, e.g., FIG. 1). The local feedback devices or resolvers 134 are embodied within or in communication with one or more components of the system 100, such as the gantry 102, the nozzle 104, the radiation source 106, the aperture 110, the imager(s) 112, patient positioner 114, patient pod 116, and/or world 132. The local feedback devices 134 provide independent position information relating to the associated component of the system 100. In various embodiments, the local feedback devices 134 comprise rotary encoders, linear encoders, servos, or other position indicators that are commercially available and whose operation is well understood by one of ordinary skill in the art. The local feedback devices 134 provide independent position information that can be utilized by the system 100 in addition to the information provided by the external measurement devices 124 to more accurately position the patient. The system 100 also comprises, in this embodiment, a precision patient alignment system 200 which employs the location information provided in state 160 for the object(s). As illustrated in FIG. 8, the patient alignment system 200 comprises a command and control module 202 communicating with a 6D system 204, a patient registration module 206, data files 210, a motion control module 212, a safety module 214, and a user interface 216. The patient alignment system 200 employs location information provided by the 6D system 204 to more accurately register the patient and move the nozzle 104 and the patient positioner 114 to achieve a desired treatment pose as indicated by the prescription for the patient provided by the data files 210. In this embodiment, the 6D system 204 receives position data from the external measurement devices 124 and from the resolvers 134 relating to the current location of the nozzle 104, the aperture 110, the imager 112, the patient positioner 114, and patient pod 116, as well as the location of one or more fixed landmarks 132 indicated in FIG. 9 as the world 132. The fixed landmarks, or world, 132 provide a non-moving origin or frame of reference to facilitate determination of the position of the moving components of the radiation therapy system 100. This location information is provided to a primary 6D position measurement system 220 which then uses the observed data from the external measurement devices 124 and resolvers 134 to calculate position and orientation coordinates of these five components and origin in a first reference frame. This position information is provided to a 6D coordination module 222 which comprises a coordinate transform module 224 and an arbitration module 226. The coordinate transform module 224 communicates with other modules of the patient alignment system 200, such as the command and control module 202 and the motion control with path planning and collision avoidance module 212. Depending on the stage of the patient registration and therapy delivery process, other modules of the patient alignment system 200 can submit calls to the 6D system 204 for a position request of the current configuration of the radiation therapy system 100. Other modules of the patient alignment system 200 can also provide calls to the 6D system 204 such as a coordinate transform request. Such a request typically will include submission of location data in a given reference frame, an indication of the reference frame in which the data is submitted and a desired frame of reference which the calling module wishes to have the position data transformed into. This coordinate transform request is submitted to the coordinate transform module 224 which performs the appropriate calculations upon the submitted data in the given reference frame and transforms the data into the desired frame of reference and returns this to the calling module of the patient alignment system 200. For example, the radiation therapy system 100 may determine that movement of the patient positioner 114 is indicated to correctly register the patient. For example, a translation of plus 2 mm along an x-axis, minus 1.5 mm along a y-axis, no change along a z-axis, and a positive 1° rotation about a vertical axis is indicated. This data would be submitted to the coordinate transform module 224 which would then operate upon the data to return corresponding movement commands to the patient positioner 114. The exact coordinate transformations will vary in specific implementations of the system 100 depending, for example, on the exact configuration and dimensions of the patient positioner 114 and the relative position of the patient positioner 114 with respect to other components of the system 100. However, such coordinate transforms can be readily determined by one of ordinary skill in the art for a particular application. The arbitration module 226 assists in operation of the motion control module 212 by providing specific object position information upon receipt of a position request. A secondary position measurement system 230 provides an alternative or backup position measurement function for the various components of the radiation therapy system 100. In one embodiment, the secondary position measurement system 230 comprises a conventional positioning functionality employing predicted position information based on an initial position and commanded moves. In one embodiment, the primary position measurement system 220 receives information from the external measurement devices 124 and the secondary position measurement system 230 receives independent position information from the resolvers 134. It will generally be preferred that the 6D measurement system 220 operate as the primary positioning system for the previously described advantages of positioning accuracy and speed. FIG. 10 illustrates in greater detail the patient registration module 206 of the patient alignment system 200. As previously described, the 6D system 204 obtains location measurements of various components of the radiation therapy system 100, including the table or patient pod 116 and the nozzle 104 and determines position coordinates of these various components and presents them in a desired frame of reference. The data files 210 provide information relating to the patient's treatment prescription, including the treatment plan and CT data previously obtained at a planning or prescription session. This patient's data can be configured by a data converter 232 to present the data in a preferred format. The imager 112 also provides location information to the 6D system 204 as well as to an image capture module 236. The image capture module 236 receives raw image data from the imager 112 and processes this data, such as with filtering, exposure correction, scaling, and cropping to provide corrected image data to a registration algorithm 241. In this embodiment, the CT data undergoes an intermediate processing step via a transgraph creation module 234 to transform the CT data into transgraphs which are provided to the registration algorithm 241. The transgraphs are an intermediate data representation and increase the speed of generation of DRRs. The registration algorithm 241 uses the transgraphs, the treatment plan, the current object position data provided by the 6D system 204 and the corrected image data from the imager(s) 112 to determine a registered pose which information is provided to the command and control module 202. The registration algorithm 241 attempts to match either as closely as possible or to within a designated tolerance the corrected image data from the imager 112 with an appropriate DRR to establish a desired pose or to register the patient. The command and control module 202 can evaluate the current registered pose and provide commands or requests to induce movement of one or more of the components of the radiation therapy system 100 to achieve this desired pose. Additional details for a suitable registration algorithm may be found in the published doctoral dissertation of David A. LaRose of May 2001 submitted to Carnegie Mellon University entitled “Iterative X-ray/CT Registration Using Accelerated Volume Rendering” which is incorporated herein in its entirety by reference. FIGS. 11-13 illustrate embodiments with which the system 100 performs this movement. FIG. 11 illustrates that the command and control module 202 has provided a call for movement of one or more of the components of the radiation therapy system 100. In state 238, the motion control module 212 retrieves a current position configuration from the 6D system 204 and provides this with the newly requested position configuration to a path planning module 240. The path planning module 240 comprises a library of three-dimensional model data which represent position envelopes defined by possible movement of the various components of the radiation therapy system 100. For example, as previously described, the imager 112 is retractable and a 3D model data module 242 indicates the envelope or volume in space through which the imager 112 can move depending on its present and end locations. The path planning module 240 also comprises an object movement simulator 244 which receives data from the 3D model data module 242 and can calculate movement simulations for the various components of the radiation therapy system 100 based upon this data. This object movement simulation module 244 preferably works in concert with a collision avoidance module 270 as illustrated in FIG. 12. FIG. 12 again illustrates one embodiment of the operation of the 6D system 204 which in this embodiment obtains location measurements of the aperture 110, imager 112, nozzle 104, patient positioner and patient pod 114 and 116 as well as the fixed landmarks or world 132. FIG. 12 also illustrates that, in this embodiment, local feedback is gathered from resolvers 134 corresponding to the patient positioner 114, the nozzle 104, the imager 112, and the angle of the gantry 102. This position information is provided to the collision avoidance module 270 which gathers the object information in an object position data library 272. This object data is provided to a decision module 274 which evaluates whether the data is verifiable. In certain embodiments, the evaluation of the module 274 can investigate possible inconsistencies or conflicts with the object position data from the library 272 such as out-of-range data or data which indicates, for example, that multiple objects are occupying the same location. If a conflict or out-of-range condition is determined, e.g., the result of the termination module 274 is negative, a system halt is indicated in state 284 to inhibit further movement of components of the radiation therapy system 100 and further proceeds to a fault recovery state 286 where appropriate measures are taken to recover or correct the fault or faults. Upon completion of the fault recovery state 286, a reset state 290 is performed followed by a return to the data retrieval of the object position data library in module 272. If the evaluation of state 274 is affirmative, a state 276 follows where the collision avoidance module 270 calculates relative distances along current and projected trajectories and provides this calculated information to an evaluation state 280 which determines whether one or more of the objects or components of the radiation therapy system 100 are too close. If the evaluation of stage 280 is negative, e.g., that the current locations and projected trajectories do not present a collision hazard, a sleep or pause state 282 follows during which movement of the one or more components of the radiation therapy system 100 is allowed to continue as indicated and proceeds to a recursive sequence through modules 272, 274, 276, 280, and 282 as indicated. However, if the results of the evaluation state 280 are affirmative, e.g., that either one or more of the objects are too close or that their projected trajectories would bring them into collision, the system halt of state 284 is implemented with the fault recovery and reset states 286 and 290, following as previously described. Thus, the collision avoidance module 270 allows the radiation therapy system 100 to proactively evaluate both current and projected locations and movement trajectories of movable components of the system 100 to mitigate possible collisions before they occur or are even initiated. This is advantageous over systems employing motion stops triggered, for example, by contact switches which halt motion upon activation of stop or contact switches, which by themselves may be inadequate to prevent damage to the moving components which can be relatively large and massive having significant inertia, or to prevent injury to a user or patient of the system. Assuming that the object movement simulation module 244 as cooperating with the collision avoidance module 270 indicates that the indicated movements will not pose a collision risk, the actual movement commands are forwarded to a motion sequence coordinator module 246 which evaluates the indicated movement vectors of the one or more components of the radiation therapy system 100 and sequences these movements via, in this embodiment, five translation modules. In particular, the translation modules 250, 252, 254, 260, and 262 translate indicated movement vectors from a provided reference frame to a command reference frame appropriate to the patient positioner 114, the gantry 102, the x-ray source 130, the imager 112, and the nozzle 104, respectively. As previously mentioned, the various moveable components of the radiation therapy system 100 can assume different dimensions and be subject to different control parameters and the translation modules 250, 252, 254, 260, and 262 interrelate or translate a motion vector in a first frame of reference into the appropriate reference frame for the corresponding component of the radiation therapy system 100. For example, in this embodiment the gantry 102 is capable of clockwise and counterclockwise rotation about an axis whereas the patient positioner 114 is positionable in six degrees of translational and rotational movement freedom and thus operates under a different frame of reference for movement commands as compared to the gantry 102. By having the availability of externally measured location information for the various components of the radiation therapy system 100, the motion sequence coordinator module 246 can efficiently plan the movement of these components in a straightforward, efficient and safe manner. FIG. 14 illustrates a workflow or method 300 of one embodiment of operation of the radiation therapy system 100 as provided with the patient alignment system 200. From a start state 302, follows an identification state 304 wherein the particular patient and treatment portal to be provided is identified. This is followed by a treatment prescription retrieval state 306 and the identification and treatment prescription retrieval of states 304 and 306 can be performed via the user interface 216 and accessing the data files of module 210. The patient is then moved to an imaging position in state 310 by entering into the patient pod 116 and actuation of the patient positioner 114 to position the patient pod 116 securing the patient in the approximate position for imaging. The gantry 102, imager(s) 112, and radiation source(s) 130 are also moved to an imaging position in state 312 and in state 314 the x-ray imaging axis parameters are determined as previously described via the 6D system 204 employing the external measurement devices 124, cooperating markers 122, and resolvers 134. In state 316, a radiographic image of the patient is captured by the imager 112 and corrections can be applied as needed as previously described by the module 236. In this embodiment, two imagers 112 and corresponding x-ray sources 130 are arranged substantially perpendicularly to each other. Thus, two independent radiographic images are obtained from orthogonal perspectives. This aspect provides more complete radiographic image information than from a single perspective. It will also be appreciated that in certain embodiments, multiple imaging of states 316 can be performed for additional data. An evaluation is performed in state 320 to determine whether the radiographic image acquisition process is complete and the determination of this decision results either in the negative case with continuation of the movement of state 312, the determination of state 314 and the capture of state 316 as indicated or, when affirmative, followed by state 322. In state 322, external measurements are performed by the 6D system 204 as previously described to determine the relative positions and orientations of the various components of the radiation therapy system 100 via the patient registration module 206 as previously described. In state 324, motion computations are made as indicated to properly align the patient in the desired pose. While not necessarily required in each instance of treatment delivery, this embodiment illustrates that in state 326 some degree of gantry 102 movement is indicated to position the gantry 102 in a treatment position as well as movement of the patient, such as via the patient positioner 114 in state 330 to position the patient in the indicated pose. Following these movements, state 332 again employs the 6D system 204 to externally measure and in state 334 to compute and analyze the measured position to determine in state 336 whether the desired patient pose has been achieved within the desired tolerance. If adequately accurate registration and positioning of the patient has not yet been achieved, state 340 follows where a correction vector is computed and transformed into the appropriate frame of reference for further movement of the gantry 102 and/or patient positioner 114. If the decision of state 336 is affirmative, e.g., that the patient has been satisfactorily positioned in the desired pose, the radiation therapy fraction is enabled in state 342 in accordance with the patient's prescription. For certain patient prescriptions, it will be understood that the treatment session may indicate multiple treatment fractions, such as treatment from a plurality of orientations and that appropriate portions of the method 300 may be iteratively repeated for multiple prescribed treatment fractions. However, for simplicity of illustration, a single iteration is illustrated in FIG. 14. Thus, following the treatment delivery of state 342, a finished state 344 follows which may comprise the completion of treatment for that patient for the day or for a given series of treatments. Thus, the radiation therapy system 100 with the patient alignment system 200, by directly measuring movable components of the system 100, employs a measured feedback to more accurately determine and control the positioning of these various components. A particular advantage of the system 100 is that the patient can be more accurately registered at a treatment delivery session than is possible with known systems and without an iterative sequence of radiographic imaging, repositioning of the patient, and subsequent radiographic imaging and data analysis. This offers the significant advantage both of more accurately delivering the therapeutic radiation, significantly decreasing the latency of the registration, imaging and positioning processes and thus increasing the possible patient throughput as well as reducing the exposure of the patient to x-ray radiation during radiographic imaging by reducing the need for multiple x-ray exposures during a treatment session. Although the preferred embodiments of the present invention have shown, described and pointed out the fundamental novel features of the invention as applied to those embodiments, it will be understood that various omissions, substitutions and changes in the form of the detail of the device illustrated may be made by those skilled in the art without departing from the spirit of the present invention. Consequently, the scope of the invention should not be limited to the foregoing description but is to be defined by the appended claims. |
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abstract | The present invention relates to a multi-core fuel rod for research reactor and, more particularly, to a multi-core fuel rod for research reactor in which monolithic fuel cores made of uranium-molybdenum alloy are disposed in an aluminum matrix in a multi-core form. The multi-core fuel rod in accordance with the present invention provides a minimized contact surface area between nuclear fuel and aluminum, and reduces the formation of pores and swelling by restraining formation of reaction layer to avoid excessive reaction between the fuel and aluminum. Therefore, improved stability of nuclear fuel can be obtained by minimizing temperature rise as well as achieving high density and thermal conductivity of the fuel. |
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summary | ||
abstract | A two dimensional (2D) collimator assembly and a detector system employing a 2D collimator assembly. More specifically, a collimator assembly is provided, having elements extending in the x and z-planes of a detector system. The 2D collimator assembly includes a number of blades arranged in parallel. Each of the blades includes fins extending from one or both sides of the body of the blade. The fins are coupled to each adjacent array to form the 2D collimator assembly. |
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description | 1. Field of the Invention The present invention relates to a reactivity control rod for a core, a core of a nuclear, a nuclear reactor and a nuclear power plant. More particularly, the present invention relates to a reactivity control rod for a core, which can elongate the lifetime of the core, a core of a nuclear reactor composed of the reactivity control rod, which can have a long lifetime, a nuclear reactor which is cooled by a liquid metal and is able to reduce scattering of the liquid metal so as to be made into a small size thereof and a nuclear power plant which comprises the nuclear reactor. 2. Description of the Related Art A conventional liquid metal cooled nuclear reactor with a small size, that is, a fast reactor is disclosed in U.S. Pat. No. 5,420,897. Moreover, a conventional fast reactor has a structure for moving a neutron reflector in a vertical direction so as to adjust (control) a leakage of neutron from the core thereof, thus to compensate a change of reactivity of the core due to a burn-up (combustion) thereof. In the aforesaid conventional liquid metal cooled nuclear reactors, an intermediate heat exchanger is arranged in a reactor vessel. A primary coolant performs the heat exchanging operation with a secondary coolant in the intermediate heat exchanger, and the exchanged secondary coolant is circulated to a steam generator arranged outside the reactor vessel so as to generate a steam. Namely, the conventional liquid metal cooled nuclear reactor has a structure of requiring a steam generator for generating a steam, an electromagnetic pump for circulating a secondary coolant between the reactor vessel and the steam generator, and piping equipments connecting them. An activated liquid metal such as sodium is used as each of the coolants. For this reason, the reactor vessel and a facility using the liquid metal arranged around the reactor vessel have complicated structures, so that there is the possibility that an auxiliary facility is required in preparation for a leakage of the activated liquid metal, fire caused thereby or the like. Moreover, in the conventional liquid metal cooled nuclear reactor, the liquid metal which is easily activated, such as sodium is used as the coolant. That is, in the steam generator, the liquid metal which is easily activated reacts to water to generate a steam. For this reason, in cases where a water leakage occurs in a heating tube of the steam generator, it is difficult to avoid an occurrence of an accident caused by the reaction between the sodium and the leaked water. The reaction between the sodium and the leaked water causes a reaction product, so that, in order to prevent the reaction product from directly being radiated, a secondary cooling system facility must be required. In addition, a facility for housing the reaction product must be required so that there is the possibility that the reactor system, as a whole, is made into a large size thereof, and that the cost of manufacturing the reactor system is made to be increased. Furthermore, the electromagnetic pump is arranged in a liquid metal; however, it is coaxially arranged in series on a downstream side (lower side) of the intermediate heat exchanger in view of a heat resistant characteristic of a large-sized conductive coil of the electromagnetic pump or the like. On the other hand, each of tube plates arranged above and below the intermediate heat exchanger has a structure which is easy to receive a thermal stress, and an enlargement of its diameter causes an increase of the thermal stress so that it is taken into consideration to prevent each of the tube plates from being made into a large size thereof. As described above, in the conventional liquid metal cooled reactor, the intermediate heat exchanger and the electromagnetic pump are vertically arranged in series; for this reason, the reactor is made into a large size thereof in its height direction (in its axial direction). The reactor with a large size in its axial direction has a structure which is easily oscillated, thereby making it unstable. On the other hand, in a conventional neutron reflector migration type of fast reactor, when elongating the lifetime of the core thereof, it must be necessary to make long the length of fuel assembly in the core. That is, according to the progress of combustion of the fuel assembly, a reactivity of the fuel assembly becomes negative. Therefore, in order to offset the negative reactivity, a neutron reflector is left up from a lower portion of the core to cover the height thereof so as to improve the ability of reflecting neutron, thereby increasing a positive reactivity of the neutron reflector, so that a reactivity of the whole core of the reactor needs to be set to 0; that is, it is necessary to make the reactor operate so as to keep a combustion in a critical state. Thus, in order to elongate an operating period of the reactor, a fuel length of the fuel assembly must be made long. Furthermore, in cases where the fuel length of the fuel assembly is made long, the reactor vessel of the reactor becomes long as a whole; as a result, there is the possibility of deteriorating the economics of the reactor. Furthermore, there are problems of causing a change of reactivity by deformation of the core in the lifetime thereof the core, an increase of pullout force of the fuel assembly. The present invention is made in view of the aforesaid problems in the related art. Accordingly, it is an object of the present invention to provide a nuclear reactor, which is capable of limiting a space for housing a liquid metal used as a coolant into an inside of a reactor vessel thereof so as to prevent scattering of the coolant to the outside thereof, whereby it is possible to make simple the whole structure of the nuclear reactor with a cooling facility, and to make compact the whole structure thereof, and to provide a nuclear power plant comprising the nuclear reactor. In order to achieve such object, according to one aspect of the present invention, there is provided a nuclear reactor in which a primary coolant is contained, including: a core composed of nuclear fuel, the coolant moving upwardly from the core by an operation thereof; an annular steam generator arranged in an upper side of the core into which the upwardly moving coolant flows and adapted to transfer heat in the coolant into water therein to generate a steam; a passage structure that defines a coolant passage for the coolant to an outside of the core, the heat-transferred coolant in the annular steam generator flowing downwardly in the coolant passage so as to flow into the core, thereby moving upwardly; and a reactor vessel arranged to surround the coolant passage so as to contain the core, the annular steam generator and the coolant passage therein. In order to achieve such object, according to another aspect of the present invention, there is provided a nuclear power plant comprising a nuclear reactor in which a coolant is contained, the nuclear reactor including: a core composed of nuclear fuel, the coolant moving upwardly from the core by an operation thereof; an annular steam generator having a plurality of heat transfer tubes and arranged in an upper side of the core into which the upwardly moving coolant flows, the annular steam generator transferring heat in the coolant with water in the heat transfer tubes to generate a steam; a passage structure that defines a coolant passage for the coolant to an outside of the core, the heat-transferred coolant in the annular steam generator flowing downwardly in the coolant passage so as to flow into the core, thereby moving upwardly; and a reactor vessel arranged to surround the coolant passage so as to contain the core, the annular steam generator and the passage means therein; a feed water branch pipe connecting to corresponding to heat transfer tubes; a steam branch pipe connecting to corresponding to heat transfer tubes, the feed water branch pipe and the steam branch pipe independently penetrating through a reactor container facility; a first feed water pipe; a steam pipe, the feed water branch pipe and the steam branch pipe being connected to the first feed water pipe and the steam pipe outside the reactor container facility, respectively; a steam bypass pipe branching from the steam branch pipe and provided with a steam separator having a bottom portion; an air conditioner provided for the steam separator via a steam facility pipe thereof; and a second feed water pipe with a feedwater pump, the bottom portion of the steam separator being connected through the second feed water pipe to the feed water branch pipe. In order to achieve such object, according to further aspect of the present invention, there is provided a reactivity control rod for use in a reactor core and for controlling a reactivity therein, comprising: a tube portion; and a mixture filled in the tube portion, the mixture being made by mixing a neutron absorber that absorbs a neutron and a neutron moderator that moderates a neutron. In order to achieve such object, according to still further aspect of the present invention, there is provided a reactor core in a core barrel of a nuclear reactor, comprising: a plurality of fuel assemblies contained in the core barrel; and a mixture contained in the core barrel, the mixture being made of a neutron absorber that absorbs a neutron in the core and a neutron moderator that moderates a neutron therein so that a reactivity of the core is controlled. According to the present invention, it is possible to reduce a heat value dispersed to the outside, thereby improving a heat efficiency thereof, and to make the reactor vessel compact into a small size as a whole, thereby securely preventing a leakage of the liquid metal. Furthermore, according to the present invention, because the whole of the reactor vessel is kept at a suitable temperature, and is protected from a rapid heat transit, it is possible to secure a structural safety of the reactor, and to make an operation of the reactor for a long period. In addition, after a shutdown of the reactor, because a natural circulating force generated by heating of the core and radiation from the reactor vessel is effectively used, it is possible to stably carry out a decay heat removal operation of the reactor. Still furthermore, in particular, the shape of the reactor is miniaturized in its longitudinal direction, and therefore, it is possible to prevent a contact of the liquid metal with the water so as to make an operation of the reactor for a long period. The preferred embodiments of the present invention will be described below with reference to the accompanying drawings. In these embodiments, as one example of a nuclear reactor according to the present invention, a liquid metal cooled reactor is described A liquid metal cooled nuclear reactor of this first embodiment is generally constructed in the following manner. More specifically, the liquid metal cooled nuclear reactor schematically comprises a reactor vessel housing therein a reflector and a neutron shield, and a partition wall structure defining a coolant passage capable of utilizing a heat generated by these reflector and neutron shield as an output of the reactor. Furthermore, the liquid metal cooled nuclear reactor comprises an electromagnetic pump and a steam generator annularly arranged, and the electromagnetic pump is arranged so as to be included in a downstream side of the steam generator so that an upper portion of the electromagnetic pump and a lower portion of the steam generator is overlapped in the axial direction thereof. The liquid metal cooling facility including the partition wall structure deciding the coolant passage, the electromagnetic pump and the steam generator and the reactor core are housed in the reactor vessel so as to make small a heat value dispersed to the outside thereof, thereby improving a heat efficiency of the reactor core, thus it is capable of making compact the reactor vessel as a whole, thereby reducing the possibility of leakage can be reduced to the utmost. FIG. 1 is a view illustrating a structure of a liquid metal cooled nuclear reactor. The liquid metal cooled nuclear reactor 1 has a core 2 composed of nuclear fuel assemblies each of which is packed with a nuclear fuel, and the core 2 is formed into a substantially cylindrical shape. The outer periphery of the core 2 is surrounded by a core barrel 3 for protecting the core 2. An annular reflector 4 is arranged outside of the core barrel 3 so as to surround the core barrel 3. An inner partition wall 6 is provided outside the reflector 4. The inner partition wall 6 surrounds the outer periphery of the reflector 4 so as to define an inner wall of a coolant passage 5 of liquid metal which is used as a primary coolant. An outer partition wall 7 defining an outer wall of the coolant passage 5 is arranged outside the inner partition wall 6 at a predetermined space. In the coolant passage 5, a neutron shield 8 is disposed so as to surround the core 2. A reactor vessel 9 is provided outside the outer partition wall 7 so as to house it, and further, a guard vessel 10 for protecting the reactor vessel 9 is arranged outside the reactor vessel 9. The reflector 4 is suspended by a plurality of driving shafts (not shown) penetrating through an upper plug 11, and is supported so as to be vertically movable by a reflector driving device (not shown). The inner partition wall 6 is extended upwardly from a base plate 12 on which the core 2 is mounted so as to form the annular coolant passage 5 between it and the outer partition wall 7, in which the neutron shield 8 is disposed. In the coolant passage 5 above the disposed neutron shield 8, an electromagnetic pump 13 and a steam generator 14 are annularly arranged, and the electromagnetic pump 13 is arranged so as to be included in a downstream side of the steam generator 14 so that an upper portion of the electromagnetic pump and a lower portion of the steam generator is overlapped in the axial direction thereof. The steam generator 14 has a shell side through which the liquid metal, which is a primary coolant, flows, and a tube side including a plurality of heat transfer tubes 16 through which water, which is a secondary coolant, flow so that a heat exchange is performed via walls of the heat transfer tubes 16 in the steam generator 14. The steam generator 14 and the electromagnetic pump 13 are arranged so that a predetermined space as a part of the coolant passage 5 is formed between an inner periphery of the steam generator 14 and an outer periphery of the electromagnetic pump 13, whereby the primary coolant discharged from a lower end portion of the steam generator 14 is sucked from the upper end portion of the electromagnetic pump 13 via the formed part of the coolant passage 5. FIG. 2 is the lateral sectional view taken along a line II—II shown in FIG. 1. As shown in FIG. 2, the core 2 is formed into a shape of circle in its lateral cross section, and the core barrel 3 is provided outside the core 2. Moreover, the reflector 4 comprises several split cylindrical elements each having both end surfaces, which are annularly arranged outside the core barrel 3 with the adjacent end surfaces of the adjacent cylindrical elements jointed to each other. The inner partition wall 6 is arranged outside the reflector 4. In this case, the end surfaces of the several split cylindrical elements extend over the entire length of the reflector 4 in the longitudinal direction thereof. In this embodiment, for example, the reflector 4 is divided into six cylindrical elements which are suspended by the driving shafts (not shown) so as to be movable in the longitudinal direction thereof without interference with each other. In FIG. 2, the neutron shield 8 comprises a plurality of cylinders 21 annularly arranged to be spaced out each other, and the cylinders 21 are arranged outside the outer periphery of the inner partition wall 6. In this first embodiment, six driving shafts (not shown) suspending the six divided cylindrical elements are arranged at a position equally separated from the center axis of the reactor vessel 9. The electromagnetic pump 13 is arranged outside the driving shaft above the core 2 via the inner partition wall 6, as shown in FIG. 1. Furthermore, the steam generator 14 is arranged outside the electromagnetic pump 13 via an outer shell 22 of the electromagnetic pump 13 and the coolant passage 5 for the primary coolant. The steam generator 14 has an inner shell 23 arranged outside the outer shell 22 of the electromagnetic pump 22 so as to surround the upper portion of the inner shell 22 thereof, an outer shell 24 arranged outside the inner shell 23 at a predetermined space so as to surround the inner shell 23, and the heat transfer tubes 16 arranged between the inner shell 23 and the outer shell 24. Still furthermore, the reactor core 1 comprises an inlet nozzle 18 which is mounted on the upper side of the reactor vessel 9 so as to be penetrated in sealed state therethrough, and an outlet nozzle 19 which is mounted on the upper side of the reactor vessel 9 so as to be penetrated in sealed state therethrough. Next, the following is a description on an operation of the liquid metal cooled nuclear reactor 1 according to the first embodiment. In the liquid metal cooled nuclear reactor 1, the core 2 uses the nuclear fuel containing plutonium or the like, and, in the actual operation of the reactor 1, the nuclear fuel including the plutonium or the like is split to generate heat, and, simultaneously, to cause excessive fast neutrons to be absorbed in depleted uranium, thereby generating plutonium on an amount equally to that to be burned up. The reflector 4 reflects the neutrons irradiated from the core 2 to thereby facilitate burn-up and breeding of the nuclear fuel in the core 2. With the burn-up of the nuclear fuel, the reflector 4 is gradually moved vertically in the axial direction of the core 2, while maintaining the criticality of the nuclear fuel. According to the vertical movement of the reflector 4, a new portion of the fuel in the core 2 is then gradually burned up, and thus keeping an operation of the reactor 1 for a long period with the burn-up maintained. In the operation of the reactor 1, the reactor vessel 9 is filled with a liquid metal, which is a primary coolant, and the core 2 is cooled by the primary coolant while taking, the outside of the core 2, heat generated by the nuclear fission therein. In FIG. 1, solid line arrow “a” of a solid line shows a flowing direction of the primary coolant. As shown by these solid line arrows, the primary coolant moves downward by the operation of the electromagnetic pump 13, and then, flows downwardly through the inside of the neutron shield 8 so as to enter into the bottom portion of the reactor vessel 9. Therefore, since the primary coolant flows through the inside of the neutron shield 8, it is possible to effectively cool the neutron shield 8. Next, the primary coolant moves upwardly while flowing through the core 2 and being heated therein, and after that, flows into the shell side of the steam generator 14 at the upper portion of the reactor vessel 9. On the other hand, as shown in FIG. 3, water used as a secondary coolant flows into the tube side of the steam generator 14. In detail, the water flowing via the inlet nozzle 18 enters into downcomer tubes 16a in the heat transfer tubes 16 to flow downwardly in the axial direction thereof. Then, the water enters into a tube side, that is, heat transfer tubes (riser tubes) 16b which are arranged in layers to flow upwardly through the heat transfer tubes 16b in the axial direction thereof. Therefore, the primary coolant flows upwardly through the shell side of the steam generator 14 and the water flows upwardly through the tube side (heat transfer tubes 16b) thereof so that heat in the primary coolant is transferred into the water in the steam generator 14, and thereafter, is discharged from the lower end portion thereof. The discharged primary coolant passes through the coolant passage 5 on the lower side of the steam generator 14 to enter into the coolant passage 5 formed as the space between the inner periphery 23 of the steam generator 14 and the outer periphery 22 of the electromagnetic pump 13, thereby moving upwardly along the coolant passage 5. The primary coolant flowing out from the coolant passage 5 is sucked from the upper end portion of the electromagnetic pump 13 via the primary coolant passage 5 formed on the upper side thereof, and is again moved downwardly by the operation of the electromagnetic pump 13. On the other hand, after the water is heated by the primary coolant in the heat transfer tubes 16b of the steam generator 14, the water flowing through the outlet nozzle 19 flows out, as a steam, to the outside of the nuclear reactor 1 (reactor vessel 9) so that a thermal power that the steam has is converted into an electric power or the like. In this first embodiment, a decay heat after the shutdown of the reactor 1 is passed through the steam generator 14 via a turbine bypass system to be removed by a condenser and a natural radiation of the reactor 1. As described above, according to the liquid metal cooled nuclear reactor 1 of this first embodiment, the all elements required for cooling the core 2 by using the liquid metal, such as the core reactor barrel 3, the reflector 4, the partition wall structure having the inner partition wall 6 and the outer partition wall 7, the neutron shield 8, the electromagnetic pump 13 and the steam generator 14 are contained in the reactor vessel 9. Therefore, it is possible to make small a heat value dispersed to the outside of the reactor 1, thereby improving the heat efficiency thereof. Furthermore, according to the liquid metal cooled nuclear reactor 1 of this first embodiment, it is possible to make compact the whole size of the reactor vessel 9, thereby reducing a possibility of leaking the liquid metal from the reactor vessel 9. Still furthermore, according to the nuclear reactor 1, because the steam generator 14 is provided in the reactor vessel 9 without providing therein an intermediate heat exchanger, it is also possible to make reduce the scale of the power plant (reactor system) using the nuclear reactor 1, and to reduce the cost of manufacturing the reactor system (power plant). In addition, according to the nuclear reactor 1 according to the first embodiment, because the electromagnetic pump 13 is arranged so as to be included in the downstream side of the steam generator 14 so that the upper portion of the electromagnetic pump 13 and the lower portion of the steam generator 14 is overlapped in the axial direction thereof, as compared with the conventional liquid metal cooled reactor having the structure that the intermediate heat exchanger and the electromagnetic pump are vertically arranged in series, it is possible to make small the length of the reactor 1 in the axial direction thereof so as to prevent the reactor 1 from being oscillated, thereby making the reactor 1 stable. Next, the following is a description on the details of the liquid metal cooled nuclear reactor 1 of the first embodiment. As shown in FIG. 1, the steam generator 14 and the electromagnetic pump 13 are constructed integrally with an upper structural member 15 of the reactor 1. The upper structural member 15 is used for suspending the steam generator 14 and the electromagnetic pump 13 together. The outer shell 24 of the steam generator 14 forms an outer shroud of the structural member 15. A seal structural member 17 comprising a piston ring or the like is interposed between the upper end portion of the inner partition wall 6 and the lower end portion of the electromagnetic pump 13. The seal structural member 17 is adapted to absorb expansion and shrinkage of the liquid metal cooled nuclear reactor 1 due to heat generated thereby so as to define the coolant passage 5. Moreover, the upper structural member 15 has a structure of integrally suspending the steam generator 14 and the electromagnetic pump 13. The expansion and shrinkage of the upper structural member 15 by thermal expansion with the operation of reactor 1 is absorbed by the seal structural member 17. A structural portion for supporting the core 2 is provided at the bottom portion of the reactor vessel 9 via the base plate 12, and the expansion and shrinkage of the reactor vessel 9 and the core 2 by heat is absorbed by the seal structural member 17. As a result, it is possible to disperse a weight loaded onto the reactor vessel 9. Moreover, the upper portion of the core 2 is a hollow space so that it is possible to perform a work of exchanging the core 2 without removing the electromagnetic pump 13 and the steam generator 14. Therefore, according to the first embodiment, in addition to the effect for making compact the reactor vessel 9 into a small size, it is possible to perform the work of exchanging the core 2 without removing the electromagnetic pump 13 and the steam generator 14. Moreover, the upper structural member 15 is provided for the reactor vessel 9 so that the upper structural member 15, the electromagnetic pump 13 and the steam generator 14 are permitted to be integrally removed therefrom, making it possible to improve the transportation and the installation of the reactor vessel 9. Subsequently, the liquid metal cooled nuclear reactor 1 will be more detailedly explained below with reference to FIG. 3 to FIG. 5. FIG. 3, FIG. 4 and FIG. 5 are enlarged views of the portions B, C and D shown in FIG. 1, respectively. Especially, FIG. 3 illustrates a state of a liquid surface of the primary coolant contained in the reactor vessel 9 when the reactor 1 is operated. In FIG. 3, FIG. 4 and FIG. 5, a solid line arrow “a” shows a flowing direction of the primary coolant when the reactor 1 is operated in the first embodiment. As shown in FIG. 5, a plurality of bypass passages 26 are formed on the structural portion supporting the base plate 12 on which the core 2 is placed, at the lower portion of the reactor vessel 9. These bypass passages 26 communicate with an annular space defined between the outer partition wall 7 and the reactor vessel 9, and the upper end portion of the outer partition wall 7 is opened in the upper space of the reactor vessel 9. As shown by these arrows “a”, the primary coolant moves downward by the operation of the electromagnetic pump 13, and then, flows downwardly through the inside of the neutron shield 8 so as to enter into the bottom portion of the reactor vessel 9. Next, most of the primary coolant is moved upwardly while flowing through the core 2 and being heated therein, and after that, flows into the shell side of the steam generator 14 at the upper portion of the reactor vessel 9. On the other hand, a part of the primary coolant flows into the annular space between the reactor vessel 9 and the outer partition wall 7 via the plurality of bypass passages 26 formed on the structural portion supporting the base plate 12, as shown in FIG. 5. The primary coolant moving up via the annular space flows over the upper end portion of the outer partition wall 7 to invert thereat, as shown in FIG. 3, and then, flows into an annular space formed between the outer partition wall 7 and the outer shell 24 of the steam generator 14. Because the primary coolant is in a low temperature state before flowing into the core 2, and is moved upwardly while cooling the whole of the reactor vessel 9, it is possible to keep, by securing a flow rate of the primary coolant, a wall surface of the reactor vessel 9 at a low temperature. Therefore, according to the first embodiment, in the operation of the reactor 1, the whole of the reactor vessel 9 is maintained at a low temperature, and thereby, it is possible to secure a structural safety of the reactor vessel 9, and to make an operation for a long period while reducing a possibility of leaking a liquid metal. Next, a decay heat removal operation will be described below with reference to FIG. 6 to FIG. 8. FIG. 6, FIG. 7 and FIG. 8 are correspondent to FIG. 3, FIG. 4 and FIG. 5, respectively. In FIG. 3, FIG. 4 and FIG. 5, the solid line arrows “a” show flowing directions of the primary coolant when removing a decay heat in this first embodiment, and broken line arrows “b” show flowing directions of air which flows in the reactor 1 through an inlet port to flow out from an outlet port. As shown in FIG. 6, the steam generator 14 has an intermediate shell 25 arranged between the inner shell 23 and the outer shell 24 for partitioning heat exchange portions (the heat transfer tubes 16b) and downcomer portions (the downcomer tubes 16a). An upper end portion of the intermediate shell 25 is projected to have a height which is higher than a liquid surface of the primary coolant in the reactor vessel 9 in a normal operation of the reactor 1. The outer shell 24 is formed with an opening portion (outer shell opening portion) 27 at a position higher than the liquid surface of the reactor vessel 9 in the normal operation of the reactor 1. Moreover, an air duct 28 is arranged to surround the outer periphery of the guard vessel 10 at a predetermined space. Moreover, the inner shell 23 of the steam generator 14 is formed at its a predetermined portion with an opening window 23a, and the predetermined portion of the opening window 23a is lower than the position of the opening portion 27 of the intermediate shall 25. In the operation of the reactor 1 shown in FIG. 3, the primary coolant heated by the core 2 flows upwardly through the upper portion of the reactor vessel 9 into the shell side of the steam generator 14. When the primary coolant flows in the shell side thereof, a liquid surface on the shell side of the steam generator 14 is the same as the liquid surface of the reactor vessel 9 on the assumption that a pressure loss of the opening window 23a may be ignored. The upper end portion of the intermediate shell 25 is projected to have the height which is higher than the liquid surface of the reactor vessel 9 in the normal operation of the reactor 1 so that it is possible to prevent the primary coolant from flowing into the downcomer tubes 16a in the heat transfer tubes 16, which are formed between the intermediate shell 25 and the outer shell 24, thereby preventing a reduction of heat efficiency of the steam generator 14. On the contrary, as shown in FIG. 6, in an operation of the reactor 1 when removing a decay heat, the liquid metal used as the primary coolant is expanded in its volume by a rise of its temperature so that the primary coolant flows over from the upper end portion of the intermediate shell 25 to flow into the annular space formed between the reactor vessel 9 and the outer partition wall 7 via the opening portion 27 formed in the outer shell 24. The primary coolant flows down through the annular space formed between the reactor vessel 9 and the outer partition wall 7, while, simultaneously, making a heat exchange with the air, shown by the arrow “b”, moving upwardly through the annular space formed between the guard vessel 10 and its outer periphery of the air duct 28 via a wall surface of the reactor vessel 9 and the wall surface of the guard vessel 10. Thereafter, as shown in FIG. 8, the primary coolant flows into the bottom portion of the reactor vessel 9 via the plurality of bypass passages 26 formed at the structural portion supporting the base plate 12 for placing the core 2. The primary coolant flowing into the bottom portion of the reactor vessel 9 and having a low temperature is sucked by a natural circulating force based on the heating in the core 2 so as to flow thereinto. Therefore, in the first embodiment, in the normal operation of the reactor 1, it is possible to highly maintain the heat efficiency of the steam generator 14, and in addition to the effect, after the shutdown of the reactor 1, it is possible to perform the operation of the reactor 1 when removing the decay heat by effectively using a natural circulating force caused by a heat generated in the core 2 and a radiation from the wall surface of the reactor vessel 9. Therefore, it is able to surely perform the removal of decay heat in the reactor 1 itself, making it possible to secure a structural safety of the reactor vessel 9, to make an operation of the reactor 1 for a long period while eliminating a probability of leaking the liquid metal. FIG. 9 to FIG. 14 illustrate a second embodiment of the present invention. FIG. 9, FIG. 10 and FIG. 11 are correspondent to FIG. 3, FIG. 4 and FIG. 5, respectively, which show a state of a liquid surface and a flow of primary coolant in an operation of the reactor. FIG. 12, FIG. 13 and FIG. 14 are correspondent to FIG. 3, FIG. 4 and FIG. 5, respectively, and show an operation in the removal of decay heat. In FIG. 9 to FIG. 14, solid line arrows “a” show flowing directions of the primary coolant, and broken lines arrow “b” show flowing directions of air. A liquid metal cooled nuclear reactor 1A of this second embodiment has basically the same structure as that of the above first embodiment and therefore, the overlapping explanation is omitted with reference to FIG. 1 and FIG. 2. In this second embodiment, as shown in FIG. 11 and FIG. 14, no bypass passage of the first embodiment is formed on a structural portion supporting the base plate 12 placing the core 2 at the lower portion of the reactor vessel 9. On the other hand, as shown in FIG. 10 and FIG. 13, the outer partition wall 7 is formed with a plurality of opening portions 29 in the vicinity of an outlet bottom portion 14a of the steam generator 14. According to the above structure of the reactor 1A, in the normal operation of the reactor 1A, as shown by the arrows “a” in FIG. 9 to FIG. 11, the primary coolant moves downward by the operation of the electromagnetic pump 13, and then, flows downwardly through the inside of the neutron shield 8 so as to enter into the bottom portion of the reactor vessel 9. Next, the primary coolant moves upwardly while flowing through the core 2 and being heated therein, and after that, flows into the shell side of the steam generator 14 at the upper portion of the reactor vessel 9. On the assumption that a pressure loss is ignored at the portion of the opening window 23a in the inner shell 23 of the steam generator 14, a liquid surface of the liquid metal in the steam generator 9 is the same as that in the reactor vessel 9. On the other hand, each liquid surface of the space between the intermediate shell 25 of the steam generator 14 and the outer shell 24 thereof, the space between the outer shell 24 and the outer partition wall 7, and the space between the outer partition wall 7 and the reactor vessel 9, is as follows. More specifically, each liquid surface of these spaces is lower than the liquid surface of the reactor vessel 9 by the pressure loss on the shell side of the steam generator 14 in the normal operation, and becomes the same liquid level. Each primary coolant in the space between the intermediate shell 25 and the outer shell 24, the space between the outer shell 24 and the outer partition wall 7, and the space between the outer partition wall 7 and the reactor vessel 9, is as follows. More specifically, in the normal operation of the reactor 1A, the primary coolant in each space gets to be an equilibrium temperature state by heat balance of an heat input from the inside of the reactor vessel 9 and a heat radiation to the air side via the reactor vessel 9 and the guard vessel 10. As a result, it is possible to protect a wall surface of the reactor vessel 9 from a rapid heat transit by a change of operating mode of the reactor 1A. Moreover, in the operation of the reactor 1A while removing decay heat, as shown by the arrows “a” in FIG. 12 to FIG. 14, the primary coolant flows over from the upper end portion of the intermediate shell 25 due to a volume expansion of the liquid metal by the rise of its temperature so that the primary coolant flows into the annular space formed between the reactor vessel 9 and the outer partition wall 7 via the opening portion 27 formed in the outer shell 24. The primary coolant gets to have a high temperature by a decay heat of the core 2, and then, flows into the annular space formed between the reactor vessel 9 and the outer partition wall 7, while, as shown by the arrow “b” in FIG. 14, making a heat exchange with the air moving upwardly through the annular space formed between the guard vessel 10 and the air duct 28 surrounding the outer periphery thereof via the wall surface of the reactor vessel 9 and the wall surface of the guard vessel 10. Thereafter, the primary coolant flows into the coolant passage 5 at the bottom portion of the steam generator 14 via the opening portions 29 formed in the outer partition wall 7 in the vicinity of the bottom portion on the outlet of the steam generator 14. Namely, the primary coolant flowing into the coolant passage 5 contributes mainly to the removal of decay heat in the wall surface of the reactor vessel 9 positioning on the outer peripheral portion of the steam generator 14. After passing through the coolant passage 5 at the bottom portion of the steam generator 14, the primary coolant is moved upwardly along an elongated portion of the coolant passage 5 formed as the space between the inner peripheral portion of the inner shell 23 of the steam generator 14 and the outer peripheral portion of the outer shell of the electromagnetic pump 13. Then, the primary coolant is sucked from the upper end portion of the electromagnetic pump 13 via the primary coolant passage 5 formed at the upper portion of the electromagnetic pump 13 so as to flow through the electromagnetic pump 13, thus to be guided downwardly. Furthermore, the primary coolant passing through the neutron shield 8 to flow into the bottom portion of the reactor vessel 9, which has a low temperature, is sucked by a natural circulating force based on heat generation in the core 2 to flow thereinto. As described above, according to this second embodiment, it is possible to protect the wall surface of the reactor vessel 9 from a rapid heat transit by a change of the operation mode of the reactor 1A, and to secure a structural safety of the reactor vessel 9, to make an operation of the reactor 1A for a long period while eliminating a probability of leaking the liquid metal. Incidentally, in this second embodiment, various modifications may be made. For example, the outer partition wall 7 shown in FIG. 9 to FIG. 11 may have a structure of removing the upper portion thereof from the vicinity of the bottom portion on the outlet of the steam generator 14, or may have a structure of forming no opening portion 29 of the outer partition wall 7. According to the above structures of the modifications, in the normal operation of the reactor according to the modifications, the primary coolant moves downward by the operation of the electromagnetic pump 13, and then, flows downwardly through the inside of the neutron shield 8 so as to enter into the bottom portion of the reactor vessel 9. Next, the primary coolant moves upwardly while flowing through the core 2 and being heated therein, and after that, flows into the shell side of the steam generator 14 at the upper portion of the reactor vessel 9. Each primary coolant in the space between the intermediate shell 25 and the outer shell 24, the space between the outer shell 24 and the outer partition wall 7, and the space between the outer partition wall 7 and the reactor vessel 9, gets to be an equilibrium temperature state by heat balance of an heat input from the inside of the reactor vessel 9 and a heat radiation to the air side via the reactor vessel 9 and the guard vessel 10. Furthermore, the primary coolant existing in the space between the intermediate shell 25 of the steam generator 14 and the outer shell 24 thereof merely receives an influence by the temperature of the downcomer tubes 16a, that is, a temperature of the water supplied to the downcomer tubes 16a so that the wall surface of the reactor vessel 9 is maintained at a relatively low temperature, and therefore, it is possible to protect a wall surface of the reactor vessel 9 from a rapid heat transit by a change of the operating mode of the reactor. On the other hand, in the operation of the reactor while removing decay heat, according to the modification, the primary coolant flows over from the upper end portion of the intermediate shell 25 due to a volume expansion of the liquid metal by the rise of its temperature so that the primary coolant flows into the annular space formed between the reactor vessel 9 and the outer partition wall 7 via the opening portion 27 formed in the outer shell 24. The primary coolant gets to have a high temperature by a decay heat of the core 2, and then, flows into the annular space formed between the reactor vessel 9 and the outer shell 24, while making a heat exchange with the air moving upwardly through the annular space formed between the guard vessel 10 and the air duct 28 surrounding the outer periphery thereof via the wall surface of the reactor vessel 9 and the wall surface of the guard vessel 10. After passing through the coolant passage 5 at the bottom portion of the steam generator 14, the primary coolant is moved upwardly along an elongated portion of the coolant passage 5 formed as the space between the inner peripheral portion of the inner shell 23 of the steam generator 14 and the outer peripheral portion of the outer shell of the electromagnetic pump 13. Then, the primary coolant is sucked from the upper end portion of the electromagnetic pump 13 via the primary coolant passage 5 formed at the upper portion of the electromagnetic pump 13 so as to flow through the electromagnetic pump 13, thus to be guided downwardly. Furthermore, the primary coolant passing through the neutron shield 8 to flow into the bottom portion of the reactor vessel 9, which has a low temperature, is sucked by a natural circulating force based on heat generation in the core 2 to flow thereinto. As described above, according to the modifications of the second embodiment, it is possible to reasonably protect the wall surface of the reactor vessel 9 from a rapid heat transit by a change of the operation mode of the reactor, and to secure a structural safety of the reactor vessel 9, to make an operation of the reactor for a long period while eliminating a probability of leaking the liquid metal. FIGS. 15A, 15B and FIG. 16 show a third embodiment of the present invention. These FIGS. 15A, 15B and FIG. 16 are correspondent to FIG. 3 and FIG. 4 as described before, respectively, and show a liquid surface state and a flow of primary coolant in an operation of reactor. In FIG. 15A and FIG. 16, arrows “a” show flowing directions of the primary coolant. A liquid metal cooled nuclear reactor 1B of this third embodiment basically has the same structure as that of the above first embodiment, and therefore, overlapping explanation is omitted with reference to FIG. 1 and FIG. 2. The liquid metal cooled nuclear reactor 1B of this third embodiment differs from the above first embodiment in that the steam generator 14 is provided with an opening portion 44 of the inner shell 23 of the steam generator 14, which communicates with a cover gas space 45 of the reactor vessel 9, and is located at the upper portion from the liquid surface of the reactor vessel 9. Moreover, in this third embodiment, each of the heating tubes 16 of the steam generator 14 has a double pipe structure provided with an inner tube 16S and an outer tube 16T surrounding an outer periphery of the inner tube 16S, as shown in FIG. 15B. In addition, the reactor 1B comprises a continuous leakage monitoring unit 46 that detects a leakage in both outer and inner tubes 16T and 16S. If a large-scale water leakage occurs in a liquid metal by simultaneous breakdown of the double tubes, a water vapor or bubble of the reaction product caused by contacting the liquid metal with the water is transferred to the surroundings from the leakage portion. In this case, in the heat exchange portion, a gas transferred upwardly from the leakage portion flows to a cover gas space of the steam generator 14. On the other hand, a gas transferred downwardly from there flows through each liquid surface of the space between the intermediate shell 25 and the outer shell 24 and the space between the outer shell 24 and the reactor vessel 9 to the cover gas space of the steam generator 14. At that time, the opening portion 44 of the inner shell 23 operates so that the cover gas space 45 of the reactor vessel 9 communicates with the cover gas space of the steam generator 14. Therefore, the water vapor or bubble of the reaction product by the large-scale water leakage generated in the liquid metal is all guided to the cover gas space 45 of the reactor vessel 9 through the opening portion 44. In this third embodiment, even if a large-scale water leakage occurs in the heating tube 16 of the steam generator 14, it is possible to maintain a safety of the reactor 1B without mixing the bubble into the core 2. Incidentally, in the third embodiment, partial modification may be made. For example, as shown in FIG. 16, the lower end portion 23b of the inner shell 23 of the steam generator 14 in the reactor 1C may be arranged at a position lower than the lower end portion 24a of the outer shell 24 thereof and the lower end portion 25a of the intermediate shell 25 in the primary coolant outlet portion of the steam generator 14. According to the above construction of the modification, if a large-scale water leakage occurs, because the lower end portion 23b of the steam generator inner shell 23 in the primary coolant outlet portion of the steam generator 14 is arranged at the position lower than the lower end portion 25a of the intermediate shell 25 and the lower end portion 24a of the outer shell 24, a gas transferred downwardly of water vapor or reaction product generated by the leakage selectively flows to the upper cover gas space of the steam generator 14 via each liquid surface of the space between the intermediate shell 25 and the outer shell 24 and the space between the outer shell 24 and the reactor vessel 9. Moreover, the opening portion 44 of the steam generator 23 operates so that the cover has space 45 of the reactor vessel 9 communicates with the cover gas space of the steam generator 14, whereby the water vapor or bubble of the reaction product by the large-scale water leakage generated in the liquid metal is all guided to the cover gas space 45 of the reactor vessel 9. In this modification of the third embodiment, even if a large-scale water leakage occurs in the heating tube 16 of the steam generator 14, it is possible to maintain a safety of the reactor 1C without mixing bubble into the core 2. Furthermore, in this third embodiment, another modification with a construction may be made. More specifically, as shown in FIG. 17, the reactor 1D comprises a detecting unit 47 that detects a peculiar change in flow rate generated due to a pressure rise of the shell side of the steam generator 14 by using a change in a current of the electromagnetic pump 13. In addition, the reactor comprises an operation control unit 48 that performs a control for stopping the operation of the electromagnetic pump 13 by a detected signal outputted from the detecting unit 47. In addition, the lower end portion 23b of the steam generator inner shell 23 is arranged at a position lower than the lower end portion 24a of the outer shell 24 and the lower end portion 25a of the intermediate shell 25. According to the above construction of the reactor 1D in the another modification, the following operation is carried out. That is, if a water vapor or reaction product gas is generated in the steam generator 14 by a large-scale water leakage, the pressure rise brings about a change in a flow rate of the primary coolant in the steam generator 14. The change in the flow rate of the primary coolant in the electromagnetic pump 13 is detected by the detecting unit 47 via the outlet portion of the steam generator 14 and the coolant passage 5, and then, the electromagnetic pump 13 stopped by the control of the operation unit 47 and, after that, the electromagnetic pump 13 is again operated. In this case, a gas transferred downwardly in the steam generator 14 is transferred selectively to the upward cover gas space thereof via each liquid surface of the space between the intermediate shell 25 and the outer shell 24 and the space between the outer shell 24 and the reactor vessel 9. Because the lower end portion 23b of the steam generator inner shell 23 is arranged at the position lower than the lower end portion 25a of the intermediate shell 25 and the lower end portion 24a of the outer shell 24. Moreover, the opening portion 44 of the steam generator 23 operates so that the reactor vessel 9 communicates with the cover gas space 45 of the steam generator 14. Therefore, the water vapor or bubble of the reaction product by the large-scale water leakage generated in the liquid metal is all guided to the cover gas space 45 of the reactor vessel 9 so that, even if a large-scale water leakage occurs in the heating tube 16 of the steam generator 14, it is possible to maintain a safety of the reactor 1D without mixing a bubble into the core 2. Moreover, another construction of a further modification according to the third embodiment may be made according to the present invention. For example, the outer tube 16T is arranged at a gap to the outer periphery of the inner tube 16S so that an inert gas such as helium or the like is sealed in the gap. Furthermore, in order to detect a leakage in both inner and outer tubes 16S and 16T, a continuous leakage monitoring unit such as a helium pressure gage, a moisture content concentration monitor or the like, is provided for the reactor according to the modification. According to the above construction of the reactor in the further modification, the heating tube 16 has a double tube structure, and the continuous leakage monitoring unit detects a leakage in both inner and outer tubes 16S and 16T by the inert gas such as helium or the like sealed in the gap between the inner and outer tubes 16S and 16T so that it is possible to securely prevent a contact of the water in the tubes 16S and 16T with the liquid metal of the shell side of the steam generator 14. Accordingly, with the above construction, because of preventing the water from contacting the liquid metal, it is possible to make a stable operation of the reactor for a long period. FIG. 18 illustrates a liquid metal cooled nuclear power plant according to a fourth embodiment of the present invention. According to the fourth embodiment, a liquid metal cooled nuclear reactor 1E basically has the same structure as that described in the one of the above embodiments, and therefore, only different points will be described below. In the liquid metal cooled nuclear reactor 1E of this fourth embodiment, the heat transfer tube 16 of the steam generator 14 shown in FIG. 3 has a double tube structure, as shown in FIG. 15B, and each of the heat transfer tubes 16 is arranged to be formed into a substantially helical shape in the heat exchange portion, in contrast with the above first embodiment. Moreover, an inert gas such as helium or the like is sealed in the space between the inner and outer tubes, and the reactor 1E is provided with a continuous leakage monitoring unit, such as a helium pressure gage, a moisture content concentration monitor, in order to detect a leakage in both inner and outer tubes, similarly to the further modification of the third embodiment. As shown in FIG. 18, the heat transfer tube 16 of the steam generator 14 is divided into a plurality of heating tube groups, and each heating tube group is connected so as to correspond to feed water and steam branch pipes. A feed water branch pipe 30 and a steam branch pipe 31 penetrate through a reactor container facility 32 independently from each other, and are connected with a feed water pipe 33 and a main steam pipe 34 outside the reactor container facility 32, respectively. Moreover, in the liquid metal cooled nuclear reactor 1E of this fourth embodiment, a steam separator 35 is provided via a main steam bypass pipe 37 branching from the steam branch pipe 31. The steam separator 35 is provided with an air condenser 36 via a steam auxiliary facility pipe 38. In addition, an auxiliary feed water pipe 39 and an auxiliary feed water pump 40 are provided as a return line to the feed-water side of the steam separator 35. According to the above construction of the fourth embodiment, in a normal operation of the reactor 1E, water flows into the feed water branching pipe 30 branching and separating from the feed water pipe 33. Then, the water flows into each of the heat transfer tubes 16 of the steam generator 14 in the reactor container facility 32, and thereafter, is heated in the heat exchange portion of each of the heat transfer tubes 16 so as to generate a steam. The steam generated in each of the heat transfer tubes 16 flows into the steam branching pipe 31 together with a steam of the identical heat transfer tube group, and passes through the reactor container facility 32. Thereafter, the steam joins with a steam from the steam branching pipe 31 of another group in the main steam pipe 34, and then, reaches a turbine 41. In the operation while removing decay heat after stopping the reactor 1E of the fourth embodiment, a steam heated by a decay heat of the reactor 1E flows into the steam branching pipe 31, and passes through the reactor container facility 32. Thereafter, the steam joins with a steam from the steam branching pipe 31 of another group in the main steam pipe 34, and then, reaches a condenser 42 by controlling a valve via a turbine bypass pipe 43. Further, when the steam decreases, the main steam pipe 34 and the turbine bypass pipe 43 are both isolated, and then, the heat of the steam is removed by the air condenser 36 via the steam separator 35 and the steam auxiliary facility pipe 38. In this case, the water condensed by the air condenser 36 passes through the steam separator 35, and then, is driven by the auxiliary feed water pump 40 so as to flow into the feed water branching pipe 30 via the auxiliary feed water pipe 39 and return to a feed-water side of the steam generator 14. In this manner, according to this fourth embodiment, it is possible to improve a reliability of the decay heat removal operation after a shutdown of the reactor 1E. Therefore, in the nuclear power plant of this fourth embodiment, the heating tube 16 of the steam generator 14 has a double tube structure, and the continuous leakage monitoring unit detects a leakage in both inner and outer tubes of each of the heat transfer tubes. Furthermore, the heating tubes 16 of the steam generator 14 are divided into a plurality of heating tube groups, and the auxiliary cooling air condenser 42 is arranged so as to be independently and correspondingly connected to these heating groups. Therefore, it is possible to securely prevent a contact of the liquid metal with the water, thus making a stable operation of the reactor 1E for a long period. In addition, even if a failure occurs in one heating tube group or the like, the operation of removing the decay heat of the reactor 1E after a shutdown thereof can be made by other heating tube having no failure and the auxiliary cooling air condenser 42. As a result, even in the case where the decay heat removal operation after a shutdown of the reactor 1E is not made by the auxiliary cooling air condenser 42, a decay heat can be removed by an operation using a heat radiation via the wall surface of the reactor vessel 9 and a natural circulating force of the primary coolant. Therefore, it is possible to secure a structural safety, and to make an operation of the reactor 1E for a long period, and further, to reduce a possibility of leakage of the liquid metal. Moreover, in this fourth embodiment, the heating tube 16 of the steam generator 14 has a double tube structure, and each heating tube 16 is formed into a helical shape in the heat exchange portion. Therefore, it is possible to arbitrarily set a dimension of the innermost layer heating tube array, and to readily provide a structure in which the electromagnetic pump 13 is housed inside the steam generator 14. In addition, each heating tube 16 has a double tube structure; therefore, it is possible to reduce a chance of contacting the water in the pipe with the liquid metal on the shell side of the steam generator 14. Accordingly, in the liquid metal cooled nuclear reactor of this fourth embodiment and the nuclear power plant using the same reactor 1E, it is possible to miniaturize the reactor in its shape, in particular, in its longitudinal direction (axial direction). Further, it is possible to securely prevent a contact of liquid metal with water, and thus to make a stable operation of the reactor for a long period. Incidentally, in this fourth embodiment, another modifications may be made. More specifically, the heating tube 16 of the steam generator 14 has a single tube structure, and each heating tube 16 is formed into a helical shape in the heat exchange portion. Further, the primary coolant is a liquid metal made of a heavy metal such as lead, lead bismuth or the like. Furthermore, the heating tube 16 of the steam generator 14 is divided into a plurality of heating tube groups, and each heating tube group is connected so as to correspond to feed water and steam branching pipes. The feed water branch pipe 30 and the steam branch pipe 31 penetrate through a reactor container facility 32 independently from each other, and are connected with the feed water pipe 33 and the main steam pipe 34 outside the reactor container facility 32, respectively. According to the above construction, although the heating tube 16 of the steam generator 14 has a single tube structure, even if a large-scale water leakage by the breakdown of heating tube occurs in a liquid metal and a heavy metal such as lead, lead bismuth or the like contacts with water, there is no of generation of reaction product, and a steam bubble is transferred from the leakage portion to the surroundings. In this case, a specific gravity of heavy metal is about ten times as much as water; therefore, most of gas is transferred upwardly from the leakage portion, and then, is transferred to the cover gas space of the steam generator 14. If the gas is transferred downwardly, the gas is transferred to the upward cover gas space via the liquid surface of the space between the intermediate shell 25 and the outer shell 24 and the space between the outer shell 24 and the reactor vessel 9. In this case, the opening portion 44 of the inner shell 23 of the steam generator 14 operates so that the cover gas space 45 of the reactor vessel 9 communicates with the cover gas space of the steam generator 14. Therefore, a water vapor or bubble of reaction product by a large-scale water leakage generated in the liquid metal is all guided to the cover gas space 45 of the reactor vessel 9, whereby, even if a large-scale water leakage occurs in the heating tube of the steam generator, it is possible to maintain a safety of the reactor without mixing the bubble into the core 2. With reference to FIG. 19A, FIG. 19B, FIG. 20 and FIG. 21, a liquid metal cooled reactor according to the fifth embodiment of the present invention is a fast reactor 1F corresponding to the liquid metal cooled reactor 1 according to the first embodiment, and the fast reactor 1F has substantially the same structure of the reactor 1 except for the core 2A. Therefore, only the structure and operation of the reactor core 2A are described hereinafter, and other elements of the fast reactor 1F are assigned to the same numerals of the reactor 1 so that the descriptions of the elements and the operations thereof are omitted. As shown in FIG. 19A and 19A, the core 2A is composed of nuclear fuel assemblies 116 which are arranged to be formed into a substantially cylindrical shape and a reactivity control assembly 119 arranged at a center portion of the fuel assemblies 116 and adapted to control the reactivity of the core 2A. In this fifth embodiment, as shown in FIG. 19, each of the fuel assemblies 116 has a hexagonal shape in its lateral cross section and the core 2A has a diameter with approximately 80 cm and an effective length thereof with approximately 200 cm. The neutron reflector 4 outside the core 2A is composed of a structural member such as stainless steel (SUS) or graphite including a cover gas space, and has a longitudinal length of approximately 200 cm and a thickness of about 15 cm. Incidentally, these measurements of the core 2A are one example of the core 2A. The partition wall 6 is arranged outside the neutron reflector 4, and the neutron shield 21 is arranged outside the partition wall 6, and further, the reactor vessel 9 is arranged outside the neutron shield 21, as described in the first embodiment of the present invention. The reactivity control assembly 119 is mounted at the center portion of the core 2A (the fuel assemblies 116). The reactivity control assembly 119 contains a mixture made by mixing neutron moderator, for example, zirconium hydride and neutron absorber, for example gadolinium. The reactivity control assembly 119 comprises, as shown in FIG. 20, a wrapper tube 120 with a hexagonal shape in its lateral cross section, and a plurality of neutron absorber rods 123, for example, seven neutron absorber rods 123 assembled to be contained therein. Each of the neutron absorber rods 123 has a cladding tube 121 and a mixture 122 which is produced by mixing a neutron moderator and a neutron absorber and is filled therein. The volume percent ratio of the neutron moderator and the neutron absorber in the mixture 122 is X to Y, wherein the X percent is bigger than the Y percent. The cladding tube 121 is made of a structural material such as a stainless steel or the like, and the mixture 122 of the neutron moderator and the neutron absorber is a mixture of zirconium hydride and gadolinium. As the gadolinium, Gd-157, Gd-155 or other similar material are able to be used. The reactivity control assembly 119, in addition to the functions for absorbing and moderating neutrons irradiated from the fuel assemblies 116, is served as a shutdown rod for a shutdown of the core 2A, and however, the reactivity control assembly 119 is not drawn out from the core 2A in the operation thereof, which is different from a shutdown rod of the conventional core. Next, the following is a description on an operation of the fifth embodiment. FIG. 21 shows various reactivity changes in the operating period (the burn-up period) of the core2A with respect to the final state thereof shown in FIG. 19. In FIG. 21, there are shown a combustion reactivity change “a” of the fuel assembly 116, a value change “b” of the neutron reflector 4, and a reactivity change “c” of the reactivity control assembly 119. According to FIG. 21, the fuel assembly 116 burned for 30 years has a great excess reactivity change; for this reason, the value change of the neutron reflector 4 can not cancel the excess reactivity of the initial fuel (fuel assembly 116). Therefore, in the case of a core having the burn-up reactivity as shown in FIG. 21 and a long lifetime of 30 years, neutron doubling is too great in the initial core even if the reactivity of the core is controlled by only the neutron reflector 4 so that no operation of the reactor 1F is performed. That is, the initial structure of the core greatly exceeds a criticality. On the contrary, in the fifth embodiment, the reactivity control assembly 119 having a function for absorbing a neutron, for example, gadolinium is contained into the core 2A so that the excessive neutrons irradiated from the core 2A are absorbed in the gadolinium in the reactivity control assembly 119, whereby the initial excess reactivity of the fuel assembly 116 is cancelled, as shown in FIG. 20. In addition, the gadolinium in the reactivity control assembly 119 is burned to be reduced so that a burn-up reactivity of the fuel is reduced while a reactivity of the reactivity control assembly 119 itself is reduced. Therefore, it is possible to perform a control of burn-up of the core 2A for a long lifetime by a combination of the reactivity change of the neutron reflector 4 by the control of the neutron reflector 4 and the reactivity change of the reactivity control assembly 119. That is, the core 2A of the reactor 1F operates so that the excess reactivity of the fuel (fuel assembly 116) substantially equals to the sum of the negative reactivity by the neutron reflector 4 and that by the reactivity control assembly 119, whereby the excess reactivity is cancelled by the sum of the negative reactivity by the neutron reflector 4 and the reactivity control assembly 119 so as to keep critical the state of the core 2A Furthermore, in this embodiment, the reactivity control assembly 119 in the core 2A of the fast reactor 1F comprises zirconium hydride used as the neutron moderator mixed with the gadolinium used as the neutron absorber so that it is possible to effectively moderate and absorb the neutrons in the core 2A. Especially, in this embodiment, it is possible to use the reactivity control assembly 119 to the fast reactor in which neutrons in the core 2A have high energy of 1.00E+05(eV), whereas, conventionally, it is hard to use the reactivity control assembly to the fast reactor. That is, FIG. 22 is a view illustrating a neutron absorption cross section of Gd-157, and that of Gd-158. According to FIG. 22, a light water reactor operates in a thermal region wherein the neutrons in the core are thermal neutrons having the energy of, for example, 1.00E−02(eV). Therefore, when the gadolinium of Gd-157 absorbs the neutrons in the core so as to get to be the gadolinium of Gd-158, because the absorption of the Gd-158 is strongly smaller than that of the Gd-157, the Gd-158 is burned so that it is unnecessary to draw the Gd-158. However, a fast reactor operates wherein the neutrons in the core are high spectrum neutrons having the energy of, for example, 1.00E+05(eV). Therefore, in a case of containing the gadolinium of Gd-157 in the core of the fast reactor, when the gadolinium of Gd-157 absorbs the neutrons in the core so as to get to be the gadolinium of Gd-158, because the absorption of the Gd-158 is substantially as well as that of the Gd-157, the Gd-158 is hardly burned so that it must be necessary to draw the Gd-158, whereby, conventionally, it may be hard to use the gadolinium of Gd-157 to the fast reactor. However, in this fifth embodiment of the present invention, because the core 2A containing the reactivity control assembly 119 including, in addition to the gadolinium, the neutron moderator, it is able to moderate the neutrons in the core 2A so as to correspond to those in a water reactor, making it possible to use the reactivity control assembly 119 to the fast reactor 1F. Next, with reference to FIG. 23, a fast reactor 1G according to a sixth embodiment of the present invention will be described below. FIG. 23 shows principal parts of the fast reactor 1G in this sixth embodiment, and corresponds to FIG. 19A. In FIG. 23, for simplification of explanation, like reference numerals are used to designate the same parts as FIG. 19A. The fast reactor 1G of the sixth embodiment is different from the fast reactor if of the above fifth embodiment in that a neutron absorber 124 with a neutron moderator is provided above the neutron reflector 4. Conventionally, the upper portion of the neutron reflector 4 is formed into a cavity in order to improve its value. In this sixth embodiment, the neutron absorber 124 with the neutron moderator is mounted into the cavity. According to the structure, in addition to the effect of the fifth embodiment, because the neutrons irradiated from the core 2A is moderated to be absorbed in the neutron absorber 124, it is possible to give a neutron shielding function to the reactor 1G, and to simplify the upper structure of the reactor 1G. Next, a fast reactor according to a seventh embodiment of the present invention will be described below. According to this seventh embodiment, the reactivity control assembly has the structure in that the distribution of the neutron moderator in the diametrical direction of the cladding tube 121 is gradually dense toward an inside of the cladding tube 121. The fast reactor of the seventh embodiment has almost the same effects as the fifth embodiment. Besides, according to the fast reactor of this seventh embodiment, it is possible to prevent a reduction of the initial neutron absorption effect, and to provide a linear reduction of the reactivity. Therefore, according to this seventh embodiment, the reactivity is linear, and the excess reactivity change by the burn-up is linear in appearance. Therefore, it is possible to linearly carry out the burn-up control by the operation of the neutron reflector 4, and thus, to carry out the operation of the neutron reflector 4 at an approximately constant speed, thereby readily performing the burn-up control. Next, a fast reactor according to an eighth embodiment of the present invention will be described below. According to this eighth embodiment, the mixture 122 in the cladding tube 121 of the reactivity control assembly 119 is formed so that the neutron moderator and the neutron absorber are mixed to be filed in the cladding tube 121, and, in this embodiment, as the neutron moderator, graphite is used. The eighth embodiment has almost the same effects as the fifth embodiment. Besides, because of using the graphite as the neutron moderator, it is possible to improve the safety of the fast reactor under the condition of high temperature, to increase the flexibility of designing the fast reactor and to correspond to the fast reactor wherein a coolant outlet temperature thereof is made high. Next, a fast reactor according to a ninth embodiment of the present invention will be described below. In this ninth embodiment, as shown in FIG. 20 in the fifth embodiment, the neutron absorber rod 123 is produced by mounting, as the mixture 122, the neutron moderator and the neutron absorber into the cladding tube 121 by a vibration compaction process. More specifically, in the case of mixing zirconium hydride and gadolinium as the mixture 122 of the neutron moderator and the neutron absorber, both zirconium hydride and gadolinium are weighted by a predetermined amount, and thereafter, are molded like granules. These granules are gradually put from a top opening portion of the cladding tube 121 whose bottom end is sealed, to be filled therein, while vibration is applied to the cladding tube 121 by a vibrator. After vibration filling, an upper plug is attached onto the top opening portion of the cladding tube 121 to be sealed thereto, and thus, the neutron absorber rod 123 is completed. In this case, the cladding tube 121 is attached on a vibration base of the vibrator, and then, a predetermined vibration is applied the cladding tube 121 thereby. According to this eighth embodiment, it is possible to simplify a process for forming the neutron absorber rod 123 containing the neutron moderator, and to carry out a remote control in forming of the neutron absorber rod 123. Furthermore, even in the case where the neutron moderator or the neutron absorber is a dangerous material such as a radioactive material, the neutron absorber rod 123 can be readily formed. Next, a fast reactor according to a tenth embodiment of the present invention will be described below. In this tenth embodiment, the cladding tube 121 or the wrapper tube 120 shown in FIG. 20 in the fifth embodiment is provided at its inner surface with an inside coat for preventing hydrogen from being transmitted, for example, a chromium coating layer. The chromium coating layer contacts with the mixture 122 of the neutron moderator and the neutron absorber, for example, the mixture of zirconium hydride and gadolinium. According to this tenth embodiment, the reactivity control assembly 119 is provided at its inner surface with the inside coat for preventing hydrogen from being transmitted, and then, the reactivity control assembly 119 is mounted into the center portion of the core 2A as shown in FIG. 19A and FIG. 19B. According to the structure, it is possible to prevent hydrogen generated by the burn-up in the core 2A from leaking outside the reactivity control assembly 119. Other effects are the same as the above fifth embodiment. Next, a fast reactor according to an eleventh embodiment of the present invention will be described below. In this eleventh embodiment, in order to improve a neutron absorptive power of the reactivity control assembly 119, the neutron absorber rod 123 is formed with the mixture 122 made by mixing a fission product (FP) as a neutron absorber and a zirconium hydride as a neutron moderator, and the neutron absorber rod 123 is mounted in the core 2A. According to this eleventh embodiment, the fission product (FP) is used as the neutron absorber, and thereby, it is possible to effectively use a radioactive material generated by another reactor, and thus, to contribute for a reduction of fission products. Other effects are the same as the fifth embodiment. Next, a fast reactor according to a twelfth embodiment of the present invention will be described below. In this twelfth embodiment, a mixture 122 of the neutron moderator and a thermal neutron absorber, for example, zirconium hydride and gadolinium in the fifth embodiment, is filled in the fuel assembly 116 at the vicinity of the central portion of the core, and thereby improving a neutron absorptive power. According to this twelfth embodiment, the fuel assembly 116 is provided with the mixture of the neutron moderator and a thermal neutron absorber, and thereby, there is no need of mounting the reactivity control assembly 119 in the central portion of the core. Further, this serves to readily make a design of the neutron absorber rod mounted in the center of the core or a neutron absorptive channel. Incidentally, in this embodiment, the mixture 122 is filled in the fuel assembly 116 in the vicinity of the central portion of the core. However, the present invention is not limited to the structure. That is, the neutron absorber may be filled in one of the fuel assemblies 116 in the vicinity of the central portion of the core, and the neutron moderator may be filled in another one of the fuel assembles 116 which is also in the vicinity of the central portion thereof. Next, a fast reactor according to a thirteenth embodiment of the present invention will be described below. In this thirteenth embodiment, in each of the aforesaid fast reactors, a mixture of a neutron moderator and a neutron absorber, for example, zirconium hydride and gadolinium, is provided in a burnable poison assembly at the central portion of the core, and thereby, a void reactivity of the final burn-up is transferred to a positive side. The reflector control type of fast reactor of this embodiment has the same function as the fifth embodiment. In general, in the fast reactor, with the burn-up of the core, a void reactivity rises to a positive side. This means that in the final burn-up, the positive reactivity is increased by spectral hardening in the case where void is generated. However, as this embodiment, in the case of the fast reactor, which is provided with the neutron absorber rod with the neutron moderator, in the final burn-up, an absorptive effect is reduced in a small neutron energy range. For this reason, in the final burn-up, the burn-up to fission is great in a low neutron energy range as compared with a general fast reactor. As a result, in the final burn-up, no transfer to a positive reactivity is made with respect to spectral hardening by coolant void generation. Therefore, in the final burn-up, the void reactivity is hard to be transferred to the positive side, and therefore, it is possible to improve safety of the fast reactor. Next, a fast reactor according to a fourteenth embodiment of the present invention will be described below. In this fourteenth embodiment, lead or lead-bismuth alloy is used in place of sodium used as the liquid metal coolant in the fifth embodiment. Other construction is the same as the fifth embodiment. According to this fifteenth embodiment, a fast neutron is moderated so as to be absorbed in the neutron absorber, and thereby, it is possible to improve a neutron absorptive power, and to provide a fast reactor which has a high neutron breeding ratio, thereby elongating a lifetime of the core. In this embodiment, the volume percent ratios of the neutron moderator and the neutron absorber in the neutron absorber rod 123 in the reactivity control assembly 119 mounted in the core 2A are not uniformed but different according to different positions in the axial direction of the core 2A. That is, the volume percent ratio of a predetermined portion of the mixture 122 in the neutron absorber rod 123 of the reactivity control assembly 119, which has a height in the axial direction thereof corresponding to the height H1 of the core 2A, is X1 to Y1, wherein the X1 percent is bigger than the Y1 percent, and the volume percent ratio of another predetermined portion of the mixture 122 in the neutron absorber rod 123 of the reactivity control assembly 119, which has a height in the axial direction thereof corresponding to the height H2 of the plenum is X2 to Y2, wherein the X2 percent is bigger than the Y2 percent, and the X1 percent and the Y1 percent are bigger than the X2 percent and the Y2 percent, respectively. Incidentally, in the above embodiments, the primary coolant, such as the liquid metal is circulated by means of the electromagnetic pump, but the present invention is not limited to the structure. That is, the electromagnetic pump is omitted in each reactor in each embodiment of the present invention, and the primary coolant is circulated by a natural circulating force generated by, for example, the heating of the core, the radiation from the reactor vessel and the like. In this modification, it is further possible to reduce the cost of manufacturing the reactor, and because of no use of the electromagnetic pump, it is possible to improve the safety of each reactor in the present invention. Furthermore, in the fifth embodiment to the fifteenth embodiment of the present invention, as a nuclear reactor, the liquid metal cooled type of fast reactor is applied, but the present invention is not limited to the structure. That is, in the fifth embodiment to the fifteenth embodiment, as a nuclear reactor, a light water reactor is able to be applied to the present invention, which has the described system for cooling the core, and furthermore, other nuclear reactors can be applied to the present invention. While there has been described what is at present considered to be the preferred embodiments and modifications of the present invention, it will be understood that various modifications which are not described yet may be made therein, and it is intended to cover in the appended claims all such modifications as fall within the true spirit and scope of the invention. The entire contents of Japanese Patent Applications H11-375240 filed on Dec. 28, 1999 and Japanese Patent Applications 2000-049031, filed on Feb. 25, 2000 are incorporated herein by reference. |
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058728260 | claims | 1. A fuel assembly provided with a fuel rod bundle in which a plurality of nuclear fuel rods containing uranium or plutonium are arranged in a channel box enclosing the fuel rod bundle, wherein a sheet of burnable poison comprised of a metal, an alloy or an intermetallic compound is integrally joined and embedded in a zirconium based alloy plate, of which said channel box is constituted, and a plurality of said sheets of said burnable poison are arranged in the periphery of said channel box. a plurality of sheets of burnable poison, each comprised of metal, an alloy or an intermetallic compound coated with a metal having a corrosion resistance, are arranged on the walls of said channel box. burnable poison is provided on the water rod and the burnable poison is coated with metal having a corrosion resistance and a sheet of burnable poison comprised of a metal, an alloy or an intermetallic compound is integrally joined and embedded in a zirconium based alloy plate, of which said channel box is constituted, and a plurality of said sheets of said burnable poison are arranged in the periphery of said channel box. a sheet of burnable poison comprised of a metal, an alloy or an intermetallic compound is integrally joined and embedded in a zirconium based alloy plate, of which said channel box is constituted, and a plurality of said sheets of said burnable poison are positioned in the periphery of said channel box and said sheets of burnable poison are so provided as to face a region of 80% or more of the effective length of the fuel rods. a sheet of burnable poison comprised of a metal, an alloy or an intermetallic compound is embedded in a zirconium based alloy plate constituting said channel box, and is hot-rolled and repeatedly cold-rolled and annealed to integrate said sheet of said burnable poison with said zirconium based alloy plate, and a plurality of said sheets of said burnable poison are arranged in periphery of said channel box. 2. A fuel assembly provided with a fuel rod bundle in which a plurality of nuclear fuel rods containing uranium or plutonium are arranged in a channel box enclosing the fuel rod bundle, wherein 3. A fuel assembly provided with a fuel rod bundle in which a plurality of fuel rods containing uranium or plutonium are arranged in a channel box enclosing the fuel rod bundle and a water rod provided in the fuel rod bundle, wherein 4. The fuel assembly according to claim 1, wherein the burnable poison is made of metal, alloy, intermetallic compound, or ceramic. 5. The fuel assembly according to claim 1, wherein the burnable poison is made of metal, alloy, or intermetallic compound, or ceramic containing at least one of cadmium, samarium, boron, gadolinium, silver, indium, and hafnium. 6. The fuel assembly according to claim 1, wherein the burnable poison is made of metal, alloy, intermetallic compound, or ceramic obtained by adding at least one of cadmium, samarium, boron, gadolinium, silver, indium, and hafnium to zirconium or zirconium-base alloy as an alloying element and forming a solid solution containing the added element in a dispersed or supersaturated state in the form of at least one of metal, intermetallic compound, oxide, hydride, and nitride. 7. The fuel assembly according to claim 1, wherein the burnable poison is made by adding at least one of cadmium, samarium, boron, gadolinium, silver, indium, and hafnium to zirconium or zirconium-group alloy as an alloying element and forming a solid-solution containing the added element in a dispersed or supersaturated state in the form of at least one of metal, intermetallic compound, oxide, hydride, and nitride. 8. The fuel assembly according to claim 1, wherein the burnable poison is more arranged nearby the corner in the cross section of the channel box viewed from its longitudinal direction. 9. The fuel assembly according to claim 8, wherein the burnable poison is symmetrically arranged in the cross section of the channel box viewed from its longitudinal direction. 10. The fuel assembly according to claim 1, wherein the burnable poison is unevenly arranged in the longitudinal direction of the channel box. 11. The fuel, assembly according to claim 10, wherein the burnable poison is more arranged at the bottom of the channel box and less arranged at the top in the longitudinal direction of the channel box. 12. A fuel assembly provided with a fuel rod bundle in which a plurality of nuclear fuel rods containing uranium or plutonium are arranged in a channel box enclosing the fuel rod bundle, wherein 13. A fuel assembly according to claim 12, wherein said sheets of burnable poison contain 5 wt. % or more of at least one of cadmium, samarium, boron, gadolinium, silver, indium, and hafnium. 14. The fuel assembly according to claim 2, wherein the burnable poison is made of metal, alloy, intermetallic compound, or ceramic. 15. The fuel assembly according to claim 3, wherein the burnable poison is made of metal, alloy, intermetallic compound, or ceramic. 16. The fuel assembly according to claim 2, wherein the burnable poison is made of metal, alloy, or intermetallic compound, or ceramic containing at least one of cadmium, samarium, boron, gadolinium, silver, indium, and hafnium. 17. The fuel assembly according to claim 3, wherein the burnable poison is made of metal, alloy, or intermetallic compound, or ceramic containing at least one of cadmium, samarium, boron, gadolinium, silver, indium, and hafnium. 18. The fuel assembly according to claim 2, wherein the burnable poison is made of metal, alloy, intermetallic compound, or ceramic obtained by adding at least one of cadmium, samarium, boron, gadolinium, silver, indium, and hafnium to zirconium or a zirconium-base alloy as an alloying element and forming a solid solution containing the added element in a dispersed or supersaturated state in the form of at least one of metal, intermetallic compound, oxide, hydride, and nitride. 19. The fuel assembly according to claim 3, wherein the burnable poison is made of metal, alloy, intermetallic compound, or ceramic obtained by adding at least one of cadmium, samarium, boron, gadolinium, silver, indium, and hafnium to zirconium or a zirconium-base alloy as an alloying element and forming a solid solution containing the added element in a dispersed or supersaturated state in the form of at least one of metal, intermetallic compound, oxide, hydride, and nitride. 20. The fuel assembly according to claim 2, wherein the burnable poison is made by adding at least one of cadmium, samarium, boron, gadolinium, silver, indium, and hafnium to a zirconium or zirconium-group alloy as an alloying element and forming a solid-solution containing the added element in a dispersed or supersaturated state in the form of at least one of metal, intermetallic compound, oxide, hydride, and nitride. 21. The fuel assembly according to claim 3, wherein the burnable poison is made by adding at least one of cadmium, samarium, boron, gadolinium, silver, indium, and hafnium to a zirconium or zirconium-group alloy as an alloying element and forming a solid-solution containing the added element in a dispersed or supersaturated state in the form of at least one of metal, intermetallic compound, oxide, hydride, and nitride. 22. The fuel assembly according to claim 2, wherein the burnable poison is unevenly provided in the cross section of the channel box viewed from its longitudinal direction. 23. The fuel assembly according to claim 2, wherein the burnable poison is more arranged nearby the corner in the cross section of the channel box viewed from its longitudinal direction. 24. The fuel assembly according to claim 2, wherein the burnable poison is unevenly arranged in the longitudinal direction of the channel box. 25. The fuel assembly according to claim 9, wherein the burnable poison is unevenly arranged in the longitudinal direction of the channel box. 26. A fuel assembly provided with a fuel rod bundle in which a plurality of nuclear fuel rods containing uranium or plutonium are arranged in a channel box enclosing the fuel rod bundle, wherein |
claims | 1. A method of preventing separation of a feedwater sparger end bracket assembly connected to a conduit of a feedwater sparger at a sparger/bracket junction within a boiling water reactor vessel, said method comprising the steps ofvertically separating an upper clamp member of a clamp from a lower clamp member of the clamp, wherein the upper and lower clamp members each have an internal compartment dispose entirely therein;locating the upper clamp member over a top of the feedwater sparger end bracket assembly, wherein the feedwater sparger end bracket assembly includes an attachment plate, the feedwater sparger includes an end plate connected to a terminal end of the conduit and to the attachment plate, and the sparger/bracket junction is comprised of the attachment plate and the end plate connected thereto;locating the lower clamp member over a bottom of the feedwater sparger end bracket assembly;moving the upper and lower clamp members toward one another to position an upper portion of the sparger/bracket junction within a the compartment of the upper clamp member and to position a lower portion of the sparger/bracket junction within the compartment of the lower clamp member;securing the upper and lower clamp members to one another over the feedwater sparger end bracket assembly with the upper and lower portions of the sparger/bracket junction positioned respectively with the compartments of the upper and lower clamp members; andleaving the upper and lower clamp members in place in the boiling water reactor vessel to prevent separation of the feedwater sparger end bracket assembly and to prevent separation of the feedwater sparger end bracket assembly from the feedwater sparger along the sparger/bracket junction. 2. The method as recited in claim 1 wherein the upper clamp member has a lower surface, the lower clamp member has an upper surface, each of the upper and lower clamp members has an inner shoulder and an outer shoulder, each of the upper and lower clamp members has internal walls defining sides of the compartment therein, and said step of leaving includes constraining the sparger/bracket junction in a first direction between the internal walls of the compartments, constraining the feedwater sparger end bracket assembly in a second direction between the lower surface of the upper clamp member and the upper surface of the lower clamp member, and constraining the feedwater sparger end bracket assembly in a third direction between the inner shoulders and the outer shoulders of the upper and lower clamp members. 3. The method as recited in claim 2 wherein the sparger/bracket junction is further comprised of a weld connecting the attachment plate to the end plate, said step of moving includes positioning an upper portion of the attachment plate and an upper portion of the end plate together within the compartment of the upper clamp member and positioning a lower portion of the attachment plate and a lower portion of the end plate together within the compartment of the lower clamp member, and said step of constraining the sparger/bracket junction in the first direction includes preventing separation of the attachment plate from the end plate in the event of cracking of the weld. 4. The method as recited in claim 2 wherein the feedwater sparger end bracket assembly further includes a shim plate disposed between and connected to the attachment plate and the end plate, the sparger/bracket junction is further comprised of the shim plate, a first weld connecting the attachment plate to the shim plate and a second weld connecting the shim plate to the end plate, said step of moving includes positioning an upper portion of the attachment plate, an upper portion of the end plate and an upper portion of the shim plate together within the compartment of the upper clamp member and positioning a lower portion of the attachment plate, a lower portion of the end plate and a lower portion of the shim plate together within the compartment of the lower clamp member, and said step of constraining the sparger/bracket junction in the first direction includes preventing separation of the attachment plate from the shim plate in the event of cracking of the first weld and preventing separation of the shim plate from the end plate in the event of cracking of the second weld. 5. The method as recited in claim 1 wherein the upper clamp member has a bore therein, the lower clamp member has a bore therein aligned with the bore of the upper clamp member, and said step of securing includes securing the upper and lower clamp members to one another with a connector received in the aligned bores of the upper and lower clamp members. 6. The method as recited in claim 1 wherein said step of leaving includes transferring loads from the feedwater sparger to the boiling water reactor vessel via the clamp. 7. The method as recited in claim 2 wherein the upper clamp member comprises a downwardly protruding tab, the lower clamp member comprises an upwardIy protruding a tab, and said step of moving includes positioning the downwardly protruding tab of the upper clamp member between the conduit of the feedwater sparger and a wall of the boiling water reactor vessel at a location spaced from the inner shoulder of the upper clamp member and positioning the upwardly protruding tab of the lower clamp member between the conduit of the feedwater sparger and the wall of the boiling water reactor vessel at a location corresponding to the location for the tab of the upper clamp member. 8. The method as recited in claim 7 wherein said step of leaving further includes balancing loads to which the clamp is subjected. 9. The method as recited in claim 2 wherein the upper clamp member has a recessed surface offset from the lower surface thereof, the lower clamp member has a recessed surface offset from the upper surface thereof, and said step of leaving further includes constraining the conduit of the feedwater sparger between the recessed surfaces of the upper and lower clamp members. 10. The method as recited in claim 1 wherein each of the upper and lower clamp members comprises an impingement shield, said step of moving further includes positioning the impingement shields between the sparger/bracket junction and a wall of the boiling water reactor vessel, and said step of leaving further includes isolating the sparger/bracket junction from the wall of the boiling water reactor water vessel via the impingement shields of the upper and lower clamp members being disposed between the sparger/bracket junction and the wall of the boiling water reactor vessel. 11. The method as recited in claim 2 wherein said step of constraining the sparger/bracket junction in the first direction includes constraining the sparger/bracket junction in a direction horizontal to the reactor vessel, said step of constraining the feedwater sparger end bracket assembly in the second direction includes constraining the feedwater sparger end bracket assembly in a direction vertical to the reactor vessel, and said step of constraining the feedwater sparger end bracket assembly in the third direction includes constraining the feedwater sparger end bracket assembly in a direction radial to the reactor vessel. 12. The method recited in claim 3 wherein the attachment plate and end plate are planar and parallel to one another in a direction vertical to the reactor vessel, the internal walls of the upper clamp member are planar and parallel to one another in the direction vertical to the reactor vessel, the internal walls of the lower clamp member are planar and parallel to one another in the direction vertical to the reactor vessel, and said step of constraining the sparger/bracket junction in the first direction includes constraining the attachment plate and end plate between the planar internal walls of the upper and lower clamp members. 13. The method recited in claim 10 wherein said step of moving includes bringing the impingement shields into engagement with one another between the sparger/bracket junction and the wall of the reactor vessel. 14. A method of preventing separation of a feedwater sparger end bracket assembly connected to a conduit of a feedwater sparger at a sparger/bracket junction within a boiling water reactor vessel, said method comprising the steps ofseparating a one piece upper clamp member of a clamp from a one piece lower clamp member of the clamp, wherein the upper and lower clamp members each have opposed internal walls defining internal compartment therein, the feedwater sparger comprises an end plate attached to an end of the conduit, and the feedwater sparger end bracket assembly comprises an attachment plate connected to the end plate;locating the internal compartment of the upper clamp member over a top portion of the attachment plate and a top portion of the end plate;locating the internal compartment of the lower clamp member over a bottom portion of the attachment plate and a bottom portion of the end plate;moving the upper and lower clamp members toward one another to confine the top portions of the attachment plate and the end plate together between the opposed internal walls of the upper clamp member and to confine the bottom portions of the attachment plate and the end plate together between the opposed internal walls of the lower clamp member;securing the upper and lower clamp members to one another over the feedwater sparger end bracket assembly; andleaving the upper and lower clamp members in place in the boiling water reactor vessel to prevent separation of the feedwater sparger end bracket assembly and to prevent separation of the feedwater sparger end bracket assembly from the feedwater sparger. 15. The method recited in claim 14 wherein the upper clamp member further includes an internal top wall between the opposed internal walls therein, the lower clamp member further includes an internal bottom wall between the opposed internal walls therein, and said step of moving includes confining the attachment plate and end plate between the internal top wall of the upper clamp member and the internal bottom wall of the lower clamp member. 16. The method recited in claim 15 wherein the feedwater sparger end bracket assembly further comprises an upper bracket member and a lower bracket member connected to the attachment plate, the upper and lower clamp members each have an outer shoulder and an inner shoulder, and said step of moving includes confining the upper bracket member of the feedwater sparger end bracket assembly between the outer and inner shoulders of the upper clamp member and confining the lower bracket member of the feedwater sparger end bracket assembly between the outer and inner shoulders of the lower clamp member. 17. The method recited in claim 16 wherein said step of moving includes positioning the outer shoulder of the upper clamp member between the upper bracket member and a wall of the reactor vessel, and positioning the outer shoulder of the lower clamp member between the lower bracket member and the wall of the reactor vessel. 18. The method recited in claim 17 wherein the upper and lower clamp members each have a shear tab, and said step of moving includes positioning the shear tab of the upper clamp member between the conduit of the feedwater sparger and the wall of the reactor vessel with a close fit and positioning the shear tab of the lower clamp member between the conduit and the wall of the reactor vessel with a close fit. 19. The method recited in claim 18 wherein said step of leaving includes placing the feedwater sparger in equilibrium via cooperation of the outer shoulders and shear tabs with the inner shoulders. 20. The method recited in claim 14 wherein the upper and lower clamp members each have an impingement shield, and said step of moving includes bringing the impingement shield of the upper clamp member into overlapping engagement with the impingement shield of the lower clamp member between the sparger/bracket junction and a wall of the reactor vessel. |
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claims | 1. A method of manufacturing a radiation-source comprising:obtaining image data of a treatment area, the image data including surface geometry and shape data;forming a glass radiation-source, constructed from neutron-activated glass, in accordance with the image data; andcausing the glass radiation-source to emit radiation. 2. The method of claim of claim 1, wherein the glass radiation-source has a shape substantially matching at least a portion the shape of the treatment areas so as to minimize radiation in non-treatment area. 3. The method of claim 1, wherein the radiation is selected from the group consisting of alpha particles, beta minus and beta plus particles, Auger electrons, gamma-rays, and x-rays. 4. The method of claim 1, wherein the glass radiation-source is constructed from neutron-activated glass selected from the group consisting of yttrium aluminosilicate, magnesium aluminosilicateholmium-166, erbium-169, dysprosium-165, rhenium-186, rhenium-188, and yttrium-90. 5. The method of claim 1, wherein the glass radiation-source is implemented as a radioisotope encased in an encasement constructed from a material selected from the group consisting of glass forming material, metallic material, and polymeric material. 6. The method of claim 1, wherein the radioisotope is selected from the group consisting of 89Sr, 90Sr, 169Yb, 32P, 33P, 90Y, 192Ir, 25I, 131I, 103Pd, 177Lu, 149Pm, 140La, 153Sm, 186Re, 166Ho, 166Dy, 137C, 57Co, 169Er, 165Dy, 97Ru, 193mPt, 195mPt, 105Rh, 68Ni, 67Cu, 64Cu, 109Cd, 111Ag, 198Au, 199Au, 201Tl, 175Yb, 47Sc, 159Gd, 212Bi, and 77As. 7. The method of claim 6, wherein the radioisotope is implemented as a particulate radioisotope. 8. The method of claim 7, wherein the radioisotope includes neutron-activated glass. 9. The method of claim 8, wherein the neutron-activated glass is selected from the group consisting of aluminosilicate, magnesium aluminosilicate, and potassium aluminogermanate containing samarium-153, holmium-166, erbium-169, dysprosium-165, rhenium-186, rhenium-188, and yttrium-90. 10. The method of claim 1, further comprising a step of adding a radiation shielding layer from a shield material so as to form an at least partially shielded glass, radiation-source. 11. A composite radiation-source comprising:a glass radiation-source, constructed from neutron-activated glass; anda shielding material connected to at least part of the glass radiation-source. 12. The composite radiation-source of claim 11, wherein the glass radiation-source is constructed from neutron-activated glass selected from the group consisting of yttrium aluminosilicate, magnesium alumino silicateholmium-166, erbium-169, dysprosium-165, rhenium-186, rhenium-188, and yttrium-90. 13. The composite radiation-source of claim 11, wherein the glass radiation-source is implemented as a radioisotope encased in an encasement, the encasement constructed from a material selected from the group consisting of glass forming material, metallic material, and polymeric material. 14. The composite radiation-source of claim 13, wherein the radioisotope is selected from the group consisting of 89Sr, 90Sr, 169Yb, 32P, 33P, 90Y, 192Ir, 25I, 131I, 103Pd, 177Lu, 149Pm, 140La, 154S, 186Re, 188Re, 166Ho, 166Dy, 137Cs, 57Co, 169Er, 165Dy, 97Ru, 193mPt, 195mPt, 105Rh, 68Ni, 67Cu, 64Cu, 109Cd, 111Ag, 198Au, 199Au, 201Tl, 175Yb, 47Sc, 159Gd, 212Bi, and 77As. 15. The composite radiation-source of claim 14, wherein the radioisotope is implemented as a particulate radioisotope. 16. The composite radiation-source of claim 15, wherein the particulate radioisotope includes neutron-activated glass. 17. The composite radiation-source of claim 16, wherein the neutron-activated glass is selected from the group consisting of aluminosilicate, magnesium aluminosilicate, and potassium aluminogermanate containing samarium-153, holmium-166, erbium-169, dysprosium-165, rhenium-186, rhenium-188, and yttrium-90. |
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description | The present application is a continuation-in-part of co-pending U.S. patent application Ser. No. 10/803,620, filed Mar. 18, 2004 now U.S. Pat No. 7,068,748. The present invention related generally to the field of storing spent nuclear fuel, and specifically to systems and methods for storing spent nuclear fuel in ventilated vertical modules. In the operation of nuclear reactors, it is customary to remove fuel assemblies after their energy has been depleted down to a predetermined level. Upon removal, this spent nuclear fuel is still highly radioactive and produces considerable heat, requiring that great care be taken in its packaging, transporting, and storing. In order to protect the environment from radiation exposure, spent nuclear fuel is first placed in a canister. The loaded canister is then transported and stored in large cylindrical containers called casks. A transfer cask is used to transport spent nuclear fuel from location to location while a storage cask is used to store spent nuclear fuel for a determined period of time. In a typical nuclear power plant, an open empty canister is first placed in an open transfer cask. The transfer cask and empty canister are then submerged in a pool of water. Spent nuclear fuel is loaded into the canister while the canister and transfer cask remain submerged in the pool of water. Once fully loaded with spent nuclear fuel, a lid is typically placed atop the canister while in the pool. The transfer cask and canister are then removed from the pool of water, the lid of the canister is welded thereon and a lid is installed on the transfer cask. The canister is then properly dewatered and filled with inert gas. The transfer cask (which is holding the loaded canister) is then transported to a location where a storage cask is located. The loaded canister is then transferred from the transfer cask to the storage cask for long term storage. During transfer from the transfer cask to the storage cask, it is imperative that the loaded canister is not exposed to the environment. One type of storage cask is a ventilated vertical overpack “VVO”). A VVO is a massive structure made principally from steel and concrete and is used to store a canister loaded with spent nuclear fuel. VVOs stand above ground and are typically cylindrical in shape and extremely heavy, weighing over 150 tons and often having a height greater than 16 feet. VVOs typically have a flat bottom, a cylindrical body having a cavity to receive a canister of spent nuclear fuel, and a removable top lid. In using a VVO to store spent nuclear fuel, a canister loaded with spent nuclear fuel is placed in the cavity of the cylindrical body of the VVO. Because the spent nuclear fuel is still producing a considerable amount of heat when it is placed in the VVO for storage, it is necessary that this heat energy have a means to escape from the VVO cavity. This heat energy is removed from the outside surface of the canister by ventilating the VVO cavity. In ventilating the VVO cavity, cool air enters the VVO chamber through bottom ventilation ducts, flows upward past the loaded canister, and exits the VVO at an elevated temperature through top ventilation ducts. The bottom and top ventilation ducts of existing VVOs are located circumferentially near the bottom and top of the VVO's cylindrical body respectively, as illustrated in FIG. 1. While it is necessary that the VVO cavity be vented so that heat can escape from the canister, it is also imperative that the VVO provide adequate radiation shielding and that the spent nuclear fuel not be directly exposed to the external environment. The inlet duct located near the bottom of the overpack is a particularly vulnerable source of radiation exposure to security and surveillance personnel who, in order to monitor the loaded overpacks, must place themselves in close vicinity of the ducts for short durations. Additionally, when a canister loaded with spent nuclear fuel is transferred from a transfer cask to a storage VVO, the transfer cask is stacked atop the storage VVO so that the canister can be lowered into the storage VVO's cavity. Most casks are very large structures and can weigh up to 250,000 lbs. and have a height of 16 ft. or more. Stacking a transfer cask atop a storage VVO/cask requires a lot of space, a large overhead crane, and possibly a restraint system for stabilization. Often, such space is not available inside a nuclear power plant. Finally, above ground storage VVOs stand at least 16 feet above ground, thus, presenting a sizable target of attack to a terrorist. FIG. 1 illustrates a traditional prior art VVO 2. Prior art VVO 2 comprises flat bottom 17, cylindrical body 12, and lid 14. Lid 14 is secured to cylindrical body 12 by bolts 18. Bolts 18 serve to restrain separation of lid 14 from body 12 if prior art VVO 2 were to tip over. Cylindrical body 12 has top ventilation ducts 15 and bottom ventilation ducts 16. Top ventilation ducts 15 are located at or near the top of cylindrical body 12 while bottom ventilation ducts 16 are located at or near the bottom of cylindrical body 12. Both bottom ventilation ducts 16 and top ventilation ducts 15 are located around the circumference of the cylindrical body 12. The entirety of prior art VVO 2 is positioned above grade. It is an object of the present invention to provide a system and method for storing spent nuclear fuel that reduces the height of the stack assembly when a transfer cask is stacked atop a storage VVO. It is another object of the present invention to provide a system and method for storing spent nuclear fuel that requires less vertical space. Yet another object of the present invention is to provide a system and method for storing spent nuclear fuel that utilizes the radiation shielding properties of the subgrade during storage while providing adequate ventilation of the spent nuclear fuel. A further object of the present invention is to provide a system and method for storing spent nuclear fuel that provides the same or greater level of operational safeguards that are available inside a fully certified nuclear power plant structure. A still further object of the present invention is to provide a system and method for storing spent nuclear fuel that decreases the dangers presented by earthquakes and other catastrophic events and virtually eliminates the potential damage from a World Trade Center or Pentagon type of attack on the stored canister. It is also an object of the present invention to provide a system and method for storing spent nuclear fuel that allows an ergonomic transfer of the spent nuclear fuel from a transfer cask to a storage VVO. Another object of the present invention is to provide a system and method for storing spent nuclear fuel below grade. Yet another object of the present invention is to provide a system and method of storing spent nuclear fuel that reduces the amount of radiation emitted to the environment. Still another object of the present invention is to provide a system and method of storing spent nuclear fuel that affords adequate heat removal capabilities from a stored canister during flood conditions, including “smart flood” conditions. These and other objects are met by the present invention which in one aspect is a system for storing spent nuclear fuel comprising: a shell forming a cavity for receiving a canister of spent nuclear fuel, at least a portion of the shell positioned below grade; and at least one inlet ventilation duct extending from an above grade inlet to a below grade outlet at or near a bottom of the cavity; the inlet ventilation duct connected to the shell so that the cavity is hermetically sealed to ingress of below grade fluids. By providing an inlet ventilation duct that extends from above grade to the cavity at a point below grade, the radiation shielding properties of the subgrade can be utilized for the spent fuel canister without obstructing the ventilation of the canister in the cavity with ambient air. When loaded with a hot spent fuel canister, cool ambient air will enter the above grade inlet, travel through the inlet ventilation duct, and enter the cavity preferably, at or near its bottom. Heat from the spent fuel will warm the cool air causing it to rise within the cavity. The heated air will then exit the cavity via an outlet ventilation duct located in either a lid or connected to the shell. Thus, below grade storage of the spent nuclear fuel canister is facilitated while affording adequate heat ventilation for the spent fuel canister. In some embodiments, it is preferable that the system further comprise a bottom plate, the bottom plate, the shell, and the at least one inlet ventilation ducts forming an integral structure. It may also be preferred that two substantially S-shaped inlet ventilation ducts be provided. In order to prohibit the incoming cool air form being heated in the inlet ventilation ducts, some embodiments will further comprise a means for insulating the inlet ventilation ducts from the shell. It further preferred in some embodiments that a major portion of the shell be positioned below grade, and that the shell be positioned sufficiently below grade so that when a canister of spent nuclear fuel is positioned in the cavity, the entire canister is below grade. In another aspect, the invention is a method of storing spent nuclear fuel comprising: providing a below grade hole; providing a system comprising a shell forming a cavity for receiving a canister of spent nuclear fuel, at least a portion of the shell positioned below grade, and at least one inlet ventilation duct extending from an inlet to an outlet at or near a bottom of the cavity, the inlet ventilation duct connected to the shell; positioning the apparatus in the hole so that the inlet of the inlet ventilation duct is above grade and the outlet of the inlet ventilation duct into the cavity is below grade; filling the hole with engineered fill; and lowering a spent fuel canister into the cavity. In some embodiments, the method will also preferably comprise the steps of lowering the canister into the cavity until the entire canister is below grade and placing a lid on a top of the shell, the lid comprising at least one outlet ventilation duct forming a passageway from at or near the top of the cavity to an ambient atmosphere. During ventilation of the canister in this embodiment, cool air will enter the cavity via the inlet ventilation duct, the cool air will be warmed by heat from the canister, and warm air will exit the cavity via the outlet ventilation duct. It is preferred that a small clearance exist between side walls of the canister and the shell when the canister is in the cavity. In other embodiments, the method may further comprise the step of positioning the canister atop support blocks positioned on a floor of the cavity, an inlet air plenum being created between the floor of the cavity and a bottom surface of the canister, wherein an outlet air plenum exists between the lid and a top surface of the canister. In some case, it may be desirable to take affirmative measures to segregate the cool air entering the cavity via the inlet ventilation duct from the warm air exiting the cavity via the outlet ventilation duct. Referring to FIGS. 2 and 3, underground VVO 20 is illustrated according to a first embodiment of the present invention. Underground VVO 20 is a vertical, ventilated dry spent fuel storage system that is fully compatible with 100 ton and 125 ton transfer casks for spent fuel canister transfer operations. Underground VVO 20 can be modified/designed to be compatible with any size or style transfer cask. Underground VVO 20 is designed to accept spent fuel canisters for storage at an Independent Spent Fuel Storage Installation “ISFSI”) in lieu of above ground overpacks (such as prior art VVO 2 in FIG. 1). All spent fuel canister types engineered for storage in free-standing and anchored overpack models can be stored in underground VVO 20. As used herein the term “canister” broadly includes any spent fuel containment apparatus, including, without limitation, multi-purpose canisters and thermally conductive casks. For example, in some areas of the world, spent fuel is transferred and stored in metal casks having a honeycomb grid-work/basket built directly into the metal cask. Such casks and similar containment apparatus qualify as canisters, as that term is used herein, and can be used in conjunction with underground VVO 20 as discussed below Underground VVO 20 comprises body 21, base 22, and removable lid 41. Body 21 is constructed of concrete, but can be constructed of other suitable materials. Body 21 is rectangular in shape but can be any shape, such as for example, cylindrical, conical, spherical, semi-spherical, triangular, or irregular in shape. A portion of body 21 is positioned below grade so that only top portion 24 protrudes above grade level 23. Preferably, at least a major portion of the height of body 21 is positioned below grade. The exact height which top portion 24 of body 21 extends above ground level 23 can be varied greatly and will depend on a multitude of design considerations, such as canister dimensions, radioactivity levels of the spent fuel to be stored, ISFSI space limitations, geographic location considering susceptibility to missile-type and ground attacks, geographic location considering frequency of and susceptibility to natural disasters (such as earthquakes, floods, tornadoes, hurricanes, tsunamis, etc.), environmental conditions (such as temperature, precipitation levels), and/or ground water levels. Preferably, top portion 24 of body 21 is less than approximately 42 inches above ground level 23, and most preferably approximately 6 to 36 inches above ground level 23. In some embodiments, it may even be preferable that the entire height of body 21 be below grade (illustrated in FIGS. 8D and 8E). As will be discussed in more detail below, when the entire height of body is below grade, only the top surface of the body will be exposed to the ambient air above grade. Referring still to FIGS, 2 and 3, body 21 forms cylindrical cavity 26 therein (best shown in FIG. 3). While cavity 26 is cylindrical in shape, cavity 26 is not limited to any specific size, shape, and/or depth and can be designed to receive and store almost any shape of canister without departing from the spirit of the invention. While not necessary to practice the invention, it is preferred that the horizontal cross-sectional size and shape of cavity 26 be designed to generally correspond to the horizontal cross-sectional size and shape of the canister-type that is to be used in conjunction with that particular underground VVO. More specifically, it is desirable that the size and shape of cavity 26 be designed so that when a spent fuel canister (such as canister 70) is positioned in cavity 26 for storage, a small clearance exists between the outer side walls of the canister and the side walls of cavity 26. Designing cavity 26 so that a small clearance is formed between the side walls of the stored canister and the side walls of cavity 26 limits the degree the canister can move within the cavity during a catastrophic event, thereby minimizing damage to the canister and the cavity walls and prohibiting the canister from tipping over within the cavity. This small clearance also facilitates flow of the heated air during spent nuclear fuel cooling. The exact size of the clearance can be controlled/designed to achieve the desired fluid flow dynamics and heat transfer capabilities for any given situation. In some embodiments, for example, the clearance may be 1 to 3 inches. A small clearance also reduces radiation streaming. Two inlet ventilation ducts 25 are provided in body 21 for providing inlet ventilation to the bottom of cavity 26. Inlet ventilation ducts 25 are elongated substantially S-shaped passageways extending from above grade inlets 27 to below grade outlets 28. Above grade inlets 27 are located on opposing side walls of top portion 24 of body 21 and open to the ambient air above ground level 23. As use herein, the terms ambient air, ambient atmosphere, or outside atmosphere, refer to the atmosphere/air external to the underground VVO, and include the natural outside environment and spaces within buildings, tents, caves, tunnels, or other man-made or natural enclosures. Below grade outlets 28 open into cavity 26 at or near its bottom at a position below the ground level 23. Thus, inlet ventilation ducts 25 provide a passageway for the inlet of ambient air to the bottom of cavity 26, despite the bottom of cavity 26 being well below grade. Vent screens 31 (FIG. 3) are provided to cover above grade inlets 27 so that objects and other debris can not enter and block the passageways of inlet ventilation ducts 25. As a result of the elongated S-shape of inlet ventilation ducts 25, above grade inlets 27 cease to be a location of elevated dose rate that is common in free-standing above ground VVOs. While below grade outlets 28 are illustrated as being opening near the bottom of the walls of cavity 26, below grade outlets 28 can be located in the floor of cavity 26 is desired. This can be accomplished by appropriately reshaping inlet ventilation ducts 25 and forming an opening through bottom plate 38 and into cavity 26. In such an embodiment, base 22 can be considered part of the body 21 through which the inlet ventilation ducts 25 extend. Above grade inlets 27 are located in the side walls of body 21 at an elevation of about 10 inches above ground level 23. However, the elevation of above grade inlets 27 is not limiting of the present invention. The inlets 27 can be located at any desired elevation above the ground level, including level/flush therewith, as shown in FIGS. 8D and 8E. Elevating above grade inlets 27 substantially above the ground level 23 helps reduce the likelihood that rain or flood water will enter the cavity 26. It is noted that for IFSI's in flood zones, floodwater can possibly rise more than a foot above ground level and, thus, enter cavity 26 via inlet ventilation ducts 25. However, as discussed below with respect to FIG. 6, underground VVO 20 is specifically designed to deal with the worst flood conditions in a safe and effective manner. While above grade inlets 27 are preferably located in the side walls of body 21, the above grade inlets are not limited to such a location and, if desired, can be located anywhere on the body, including for example in the top surface (or any other surface) of the body. Further examples of possible locations for above grade inlets 27 on body 21 are illustrated in FIGS. 8A-8E. Referring still to FIGS. 2 and 3, inlet ventilation ducts 25 have a rectangular cross-sectional area of about 6 inches by 40 inches. However, any cross-sectional shape and/or size can be used, such as for example, round, elliptical, triangular, hexagonal, octagonal, etc. Additionally, while the shape of inlet ventilation ducts 25 is an elongated substantially S-shaped passageway, a multitude of shapes can be used that still achieve acceptable dose rates at the above grade inlets 27. For example, rather than an elongated S-shape, the inlet ventilation duct can extend from the above grade inlet to the below grade outlet in a zig-zag shape, a tilted linear shape, a general L-shape, or any angular, linear, or curved combination. The exact shape, size, and cross-sectional configuration of the inlet ventilation duct is a matter of design preference and will be dictated by such factors, such as thickness of the body of the VVO, radioactivity level of the spent fuel being stored in the cavity, temperature of the spent fuel canister, desired fluid flow dynamics through the ducts, and placement of the above grade inlet vents on the body (i.e., whether the above grade inlet vents/opening are located on the side walls of the body, its top surface, or some other surface of the body). Further examples of possible shapes for inlet ventilation ducts 25 are illustrated in FIGS. 8A-8E. Inlet ventilation ducts 25 are preferably formed by a low carbon steel liner. However, inlet ventilation ducts 25 can be made of any material or can be mere passageways formed into concrete body 21 without a lining. As best illustrated in FIG. 3, cavity 26 is formed by thick steel shell 34 and bottom plate 38. Shell 34, bottom plate 38, and inlet ventilation ducts 25 are preferably made of a metal, such as steel, preferably low carbon steel, but can be made of other materials, such as stainless steel, aluminum, aluminum-alloys, plastics, and the like. Inlet ventilation ducts 25 are seal joined to shell 34 and bottom plate 38 to form an integral/unitary structure 100 (shown in isolation in FIG. 9) that is hermetically sealed to the ingress of below grade water and other fluids. In the case of weldable metals, this seal joining may comprise welding or the use of gaskets. Thus, the only way water or other fluids can enter cavity 26 is through above grade inlets 27 or outlet ventilation ducts 42 in lid 41. As will be discussed below with respect to FIGS. 9-15, the integral structure itself is an invention and can be used to store spent nuclear fuel without the use of body 21. An appropriate preservative, such as a coal tar epoxy or the like, is applied to the exposed surfaces of shell 34, bottom plate 38, and inlet ventilation ducts 25 in order to ensure sealing, to decrease decay of the materials, and to protect against fire. A suitable coal tar epoxy is produced by Carboline Company out of St. Louis, Mo. under the tradename Bitumastic 300M. In some embodiments of the underground VVO of the present invention, a bottom plate will not be used. Concrete body 21 surrounds shell 34 and inlet ventilation ducts 25. Body 21 provides non-structural protection for shell 34 and inlet ventilation ducts 25. Insulation 37 is provided at the interface between shell 34 and concrete body 21 and at the interface between inlet ventilation ducts 25 and concrete body 21. Insulation 37 is provided to prevent excessive transmission of heat decay from spent fuel canister 70 to concrete body 21, thus maintaining the bulk temperature of the concrete within FSAR limits. Insulating shell 34 and inlet ventilation ducts 25 from concrete body 21 also serves to minimize the heat-up of the incoming cooling air before it enters cavity 26. Suitable forms of insulation include, without limitation, blankets of alumina-silica fire clay (Kaowool Blanket), oxides of alimuna and silica (Kaowool S Blanket), alumina-silica-zirconia fiber (Cerablanket), and alumina-silica-chromia (Cerachrome Blanket). Insulating inlet ventilation ducts 25 from the heat load of spent fuel in cavity 26 is very important in facilitating and maintaining adequate ventilation/cooling of the spent fuel. The insulating process can be achieved in a variety of ways, none of which are limiting of the present invention. For example, in addition to adding an insulating material to the exterior of the shell 34 and inlet ventilation ducts 25, it is also possible to insulate inlet ventilation ducts 25 by providing a gap in concrete body 21 between cavity 26 and inlet ventilation ducts 25. The gap may be filled with an inert gas or air if desired. Moreover, irrespective of the means used to provide the insulating effect, the insulating means is not limited to being positioned on the outside surfaces of shell 34 or inlet ventilation ducts 25 but can be positioned anywhere between cavity 26 and inlet ventilation ducts 25. Body 21, along with the integral steel unit formed by bottom plate 38, shell 34, and ventilation ducts 25, are placed atop base 22. Base 22 is a reinforced concrete slab designed to satisfy the load combinations of recognized industry standards, such as, without limitation, ACI-349. Base 22 is rectangular in shape but can take on any shape necessary to support body 21, such as round, elliptical, triangular, hexagonal, octagonal, irregularly shaped, etc. While using a base is preferable to achieve adequate load supporting requirements, situations can arise where using such a base may be unnecessary. Referring back to FIG. 2, underground VVO 20 has a removable ventilated lid 41. Lid 41 is positioned atop body 21, thereby substantially enclosing cavity 26 so that radiation does not escape through the top of cavity 26 when canister 70 is positioned in cavity 26. When lid 41 is placed atop body 21 and spent fuel canister 70 is positioned in cavity 26, outlet air plenum 36 is formed between the top surface of canister 70 and lid 41. Outlet air plenum 36 is preferably a minimum of 3 inches in height, but can be any desired height. The exact height will be dictated by design considerations such as desired fluid flow dynamics, canister height, VVO height, the depth of the cavity, canister heat load, etc. Lid 41 has four outlet ventilation ducts 42. Outlet ventilation ducts 42 form a passageway from the top of cavity 26 (specifically from outlet air plenum 36) to the ambient air so that heated air can escape from cavity 26. Outlet ventilation ducts 42 are horizontal passageways that extend through side wall 30 of lid 41. However, the outlet ventilation ducts can be any shape or orientation, such as vertical, L-shaped, S-shaped, angular, curved, etc. Because outlet ventilation ducts 42 are located within lid 41 itself, the total height of body 21 is minimized. Lid 41 comprises a roof 35 made of concrete. Roof 35 provides radiation shielding so that radiation does not escape from the top of cavity 26. Side wall 30 of lid 41 is an annular ring. Outlet air plenum 36 helps facilitate the removal of heated air via outlet ventilation ducts 42. In order to minimize the heated air exiting outlet ventilation ducts 42 from being siphoned back into inlet ventilation ducts 25, outlet ventilation ducts 42 are azimuthally and circumferentially separated from inlet ventilation ducts 25. Ventilated lid 41 also comprises shear ring 47. When lid 41 is placed atop body 21, shear ring 47 protrudes into cavity 26, thus, providing enormous shear resistance against lateral forces from earthquakes, impactive missiles, or other projectiles. Lid 41 is secured to body 21 with bolts (not shown) that extend therethrough. While not illustrated, it is preferable that duct photon attenuators be inserted into all of inlet ventilation ducts 25 and/or outlet ventilation ducts 42 of underground VVO 20, irrespective of shape and/or size. A suitable duct photon attenuator is described in U.S. Pat. No. 6,519,307, Bongrazio, the teachings of which are incorporated herein by reference. Referring now to FIG. 4, an embodiment of a lid 50 that can be used in underground VVO 20 is illustrated. Lid 50 contains similar design aspects as lid 41 and is illustrated to more fully disclose the aforementioned lid design aspects. Lid 50 has four horizontal outlet ventilation ducts 51 in side wall 52. Shear ring 54 is provided on the bottom of lid 50 to fit into cavity 26. Bolts 18 are used to secure lid 50 to tapped holes in the top of body 21. While the outlet ventilation ducts are illustrated as being located within the lid 50 of underground VVO 20, the present invention is not so limited. For example, outlet ventilation ducts can be located in the body of the underground VVO at a location above grade. This concept is illustrated if FIGS. 8A-8E. If the outlet ventilation ducts are located in the body of the underground VVO, the openings of the outlet ventilation ducts to the ambient air can be located in the body's side walls, on its top surface, or in any other surface. Similar to when the outlet ventilation ducts are located in the lid, the outlet ventilation ducts can take on a variety of shapes and/or configurations when located in the body of the underground VVO itself. As with the inlet ventilation ducts, the outlet ventilation ducts are preferably formed by a low carbon steel liner, but can be made of any material or can be mere passageways formed into concrete body 21 or lid 41 without a lining. In all embodiments of the present invention which have both inlet and outlet ventilation ducts, it is preferred that the outlet ventilation duct openings be azimuthally and circumferentially separated from the inlets of the inlet ventilation ducts to minimize interaction between inlet and outlet air streams. There is no limitation on the shape and style of lid used in conjunction with underground VVO 20. Referring back to FIG. 2, soil 29 surrounds body 21 for almost the entirety of its height. When spent fuel canister 70 is positioned in cavity 26, at least a major portion, if not the entirety, of canister 70 is below grade. Preferably, the entire height of canister 70 is below grade in order to take full advantage of the shielding effect of the soil 29. Thus, soil 29 provides a degree of radiation shielding for spent fuel stored in underground VVO 20 that can not be achieved in above-ground overpacks. Underground VVO 20 is unobtrusive in appearance and there is no danger of underground VVO 20 tipping over. Additionally, underground VVO 20 does not have to contend with soil-structure interaction effects that magnify the free-field acceleration and potentially challenge the stability of an above ground free-standing overpack. Referring to FIG. 6, area VI-VI of FIG. 2 is illustrated in detail. FIG. 6 illustrates design aspects that are important to ensure that underground VVO 20 can successfully withstand flood conditions without adverse impact. Support blocks 32 are provided on the bottom surface (formed by plate 38) of cavity 26 so that canister 70 can be placed thereon. Support blocks 32 are circumferentially spaced from one another (shown in FIG. 7). When canister 70 is loaded into cavity 26 for storage, the bottom surface 71 of canister 70 rests on support blocks 32, forming an inlet air plenum 33 between the bottom surface 71 of the canister 70 and the bottom surface/floor of cavity 26. Support blocks 32 are made of low carbon steel and are preferably welded to the bottom surface of the cavity 26. Other suitable materials of construction include, without limitation, reinforced-concrete, stainless steel, and other metal alloys. Support blocks 32 also serve an energy/impact absorbing function. Support blocks 32 are preferably of a honeycomb grid style, such as those manufactured by Hexcel Corp., out of California, U.S. Support blocks 32 are specifically designed so that bottom surface 71 of canister 70 is lower than top 74 of below grade outlets 28 (FIG. 2) of inlet ventilation ducts 25. Preferably, support blocks 32 are designed so that bottom surface 71 of canister 70 is about 2 to 6 inches below top 74 of below grade outlets 28. However, any desired height differential can be achieved through proper design. By supporting canister 70 in cavity 26 so that its bottom surface 71 is lower than top 74 of below grade outlets 28, underground VVO 20 will provide adequate cooling to canister 70 under even the most adverse flood condition, which is colloquially referred to as a “smart flood.” A “smart flood” is one that floods the VVO so that the water level is just high enough to block airflow though the inlet ventilation ducts 25 completely. In other words, the water level is just even with top 74 of the below grade outlets 28. However, underground VVO 20 can adequately deal with the “smart flood” condition because the bottom surface 71 of the canister 70 is situated at a height that is below top 74 of below grade outlets 28. As a result, if a “smart flood” was to occur, the bottom of the canister 70 will be in contact with (i.e. submerged in) the water. Because the heat removal efficacy of water is over 100 times that of air, a wet bottom is all that is needed to effectively remove heat and keep the canister 70 cool. The deeper the submergence of canister 70 in the water, the cooler canister 70 and its contained fuel will remain. As the water in cavity 26 is heated by the bottom of canister 70, the water evaporates, rises through cavity 26 via annular space 60, and exits cavity 26 via the outlet ventilation ducts. Thus, the canister cooling action changes from ventilation air-cooling to evaporative water cooling. In one embodiment, below grade outlets 28 of inlet ventilation ducts 25 will be 8 inches high by 40 inches wide and inlet air plenum 33 is 6 inches high. This provides a height differential of 2 inches. It should be noted that the height differential design aspect of underground VVO 20 that is detailed in FIG. 6 can also be incorporated into free-standing above ground casks and VVOs to deal with “smart flood” conditions, independent of the other features of underground VVO 20. Thus, this concept is an independent inventive aspect of the present application. When incorporated into above ground VVOs, the inlet ventilation ducts should be designed so that radiation can not escape to the surrounding environment from the inlet ventilation ducts. This is a threat because the canister will be below the inlet duct's opening into the storage cavity. In this embodiment, the inlet ventilation ducts will be shaped so that a line of sight does not exist to the canister in the storage cavity from the ambient air. For example, the inlet ventilation ducts can comprise a portion that is L-shaped, angled, S-shaped, or curved. Moreover, while the height differential design aspect of FIG. 6 is achieved using support blocks 32, it is also possible to practice this aspect of the invention without support blocks 32. In such embodiments, canister 70 will be positioned in cavity 26 and rest directly on the floor of cavity 26. However, the use of support blocks 32 is desirable because of the creation of air inlet plenum 33 and because the use of support blocks 32 helps prohibit debris and dirt from getting trapped at the bottom of cavity 26. Referring now to FIGS. 8A-8E, examples of alternative configurations of the outlet ventilation ducts and the inlet ventilation ducts in an underground VVO according to the present invention are schematically illustrated. Much of the detail, and some structure, has been omitted in FIGS. 8A-8E for simplicity with the understanding that any or all of the details discussed above with respect to underground VVO 20 can be incorporated therein. Like numbers are used to identify like parts with the exception of alphabetical suffixes being used for each embodiment. It should be noted that, in addition to the configurations of the inlet ventilation ducts and the outlet ventilation ducts illustrated in FIGS. 8A-8E, a multitude of other configurations, combinations, and modifications can be incorporated into the present invention. Some of these details are discussed above. Additionally, the outlet ventilation duct configurations of any of the illustrated embodiments can be combined with any of the illustrated inlet ventilation duct configurations, and vice versa. In all embodiments of the present invention, it is desirable that the heated air exiting the outlet ventilation ducts 42 be prohibited from being siphoned back into the inlet ventilation ducts 25 (i.e., keeping the warm outlet air stream from mixing with the cool inlet air stream). This can be accomplished by in a number of ways, including: (1) the positioning/placement of the inlets 27 on the underground VVO 20 with respect to the outlets of the outlet ventilation ducts 42; providing a plate 98 or other structure that segregates the air streams (as exemplified in FIGS. 8A and 8C-8E); and/or (3) extending the inlet ventilation ducts 25 to a position away from the outlet ventilation ducts 42. As a result of the heat emanating from canister 70, cool air from the ambient is siphoned into inlet ventilation ducts 25 and into the bottom of cavity 26. This cool air is then warmed by the heat from the spent fuel in canister 70, rises in cavity 26 via annular space 60 (FIG. 6) around canister 70, and then exits cavity 26 as heated air via outlet ventilation ducts 42 in lid 41. Referring now to FIGS. 5, ISFIs can be designed to employ any number of underground VVOs 20 (or integral structures 100) and can be expanded in number easily to meet growing needs. Although underground VVOs 20 are closely spaced, the design permits any cavity to be independently accessed by cask crawler 90 with ease. The subterranean configuration of underground VVOs 20 greatly reduce the height of the stack structures created during loading/transfer procedures where transfer cask 80 is positioned atop underground VVO 20. An embodiment of a method of using underground VVO 20 to store spent nuclear fuel canister 70 will now be discussed in relation to FIGS. 2-5. Upon being removed from a spent fuel pool and treated for dry storage, spent fuel canister 70 is positioned in transfer cask 80. Transfer cask is 80 is carried by cask crawler 90 to a desired underground VVO 20 for storage. While a cask crawler is illustrated, any suitable means of transporting transfer cask 80 to a position above underground VVO 20 can be used. For example, any suitable type of load-handling device, such as without limitation, a gantry crane, overhead crane, or other crane device can be used. In preparing the desired underground VVO 20 to receive canister 70, lid 41 is removed from body 21 so that cavity 26 is open. Cask crawler 90 positions transfer cask 80 atop underground VVO 20. After transfer cask is properly secured to the top of underground VVO 20, a bottom plate of transfer cask 80 is removed. If necessary, a suitable mating device can be used to secure the connection of transfer cask 80 to underground VVO 20 and to remove the bottom plate of transfer cask 80 to an unobtrusive position. Such mating devices are well known in the art and are often used in canister transfer procedures. Canister 70 is then lowered by cask crawler 90 from transfer cask 80 into cavity 26 of underground VVO 20 until the bottom surface of canister 70 contacts and rests atop support blocks 32, as described above. When resting on support blocks 32, a major portion of the canister's height is below grade. Most preferably, the entirety of canister 70 is below grade when in its storage position. Once canister 70 is positioned and resting in cavity 26, lid 41 is placed over cavity 26, substantially enclosing cavity 26. Lid 41 is oriented atop body 21 so that shear ring 47 protrudes into cavity 26 and outlet ventilation ducts 42 are azimuthally and circumferentially separated from inlet ventilation ducts 25 on body 21. Lid 41 is then secured to body 21 with bolts. As a result of the heat emanating from canister 70, cool air from the ambient is siphoned into inlet ventilation ducts 25 and into the bottom of cavity 26. This cool air is then warmed by the heat from the spent fuel in canister 70, rises in cavity 26 via annular space 60 (FIG. 6) around canister 70, and then exits cavity 26 as heated air via outlet ventilation ducts 42 in lid 41. Referring now to FIG. 9, an integral structure 100 for storing spent nuclear fuel is illustrated according to an embodiment of the invention. Integral structure 100 is essentially a combination of shell 34, inlet ventilation ducts 25, and bottom plate 38 of underground VVO 20 without the concrete body. Integral shell 100 can be used to store canisters of spent nuclear fuel without the addition of the concrete body. Therefore, some embodiments of the present invention will be the integral structure 100 itself. Shell 34, bottom plate 38, and inlet ventilation ducts 25 are preferably formed of a metal, such as low carbon steel. Other suitable materials include, without limitation, stainless steel, aluminum, aluminum-alloys, plastics, and the like. Inlet ventilation ducts 25, bottom plate 38, and shell 34 are seal welded at all junctures to form a unitary structure that is hermetically sealed to the ingress water and other fluids. The only way water or other fluids can enter cavity 26 is through inlets 27 or top opening 101 of shell 34. The height of shell 34 is designed so that a canister of spent fuel can be positioned within cavity 26 so as not to protrude from top opening 101. There is no limitation on the height to which shell 34 can be constructed. The exact height of shell 34 will be dictated by the height of the spent fuel canister to be stored therein, the desired depth (below grade) at which the canister is to be stored, whether the outlet ventilation ducts are in the lid or integrated into the shell 34, and/or the desired height of the outlet air plenum that is to exist during canister storage. FIGS. 10-13 illustrate a process of using integral structure 100 to store a spent fuel canister at a below grade position at an ISFSI, or other location, according to one embodiment of the present invention. It should be noted that the any of the design and/or structural details discussed above with respect to underground VVO 20 can be incorporated into integral structure 100, such as, for example, the use of vent screens, variable configurations of the inlet and outlet ducts, clearances, the use of an insulation, etc. However, in order to avoid redundancy, a discussion of these details will be omitted with the understanding that any or all of the details of underground VVO 20 are (or can be) incorporated into the storing methods and apparatus of integral structure 100, and vice versa. Referring to FIG. 10, a hole 200 is first dug into the ground 210 at a desired position within the ISFSI and at a desired depth. Once hole 200 is dug, and its bottom properly leveled, base 22 is placed at the bottom of hole 200. Base 22 is a reinforced concrete slab designed to satisfy the load combinations of recognized industry standards, such as ACI-349. However, in some embodiments, depending on the load to be supported and/or the ground characteristics, the use of a base may be unnecessary. Once base 22 is properly positioned in hole 200, integral structure 100 is lowered into the hole 200 in a vertical orientation until it rests atop base 22. Bottom plate 38 of integral structure 100 contacts and rests atop the top surface of base 22. If desired, the bottom plate 38 can be bolted or otherwise secured to the base 22 at this point to prohibit future movement of the integral structure 100 with respect to the base 22. Referring to FIG. 11, once integral structure 100 is resting atop base 22 in the vertical orientation, soil supply pipe 300 is moved into position above hole 200. Soil 301 is delivered into hole 200 exterior of integral structure 100, thereby filling hole 200 with soil 301 and burying a portion of the integral structure 100. While soil 301 is exemplified to fill hole 200, any suitable engineered fill can be used that meets environmental and shielding requirements. Other suitable engineered fills include, without limitation, gravel, crushed rock, concrete, sand, and the like. Moreover, the desired engineered fill can be supplied to the hole by any means feasible, including manually, dumping, and the like. Referring to FIG. 12, soil 301 is supplied to hole 200 until soil 301 surrounds integral structure 100 and fills hole 200 to a level where soil 301 is approximately equal to ground level 212. Soil 301 is in direct contact with the exterior surfaces of integral structure 100 that are below grade. When hole 200 is filled with soil 301, inlets 27 of inlet ventilation ducts 25 are above grade. Shell 34 also protrudes from soil 301 so that opening 101 is slightly above grade. Therefore, because integral structure 100 is hermetically sealed at all junctures, below grade liquids and soil can not enter into cavity 26 or inlet ventilation ducts 25. Support blocks 32 are provided at the bottom of cavity 26 for supporting a stored spent fuel canister. Referring to FIG. 13, once hole 200 is adequately filled with soil 301, a canister 70 of spent fuel 70 is loaded into cavity 26 of integral structure 100. The canister loading sequence is discussed in greater detail above with respect to FIG. 5. Canister 70 is lowered into cavity 26 until it rests on support blocks 32. As discussed above with respect to FIG. 6, support blocks 32 and outlets 28 of integral structure 100 are specially designed to deal with “smart flood” conditions. Canister 70 rests on support blocks 32, forming an inlet air plenum 33 between the bottom of canister 70 and the floor of cavity 26 (which in this case is bottom plate 38). When canister 70 is supported on support blocks 32, the entire height of canister 70 is below ground level 212. This maximizes use of the ground's radiation shielding capabilities. The depth at which canister 70 is below ground level 212 can be varied by increasing or decreasing the depth of hole 200. Once canister 70 is supported in cavity 26, lid 41 is placed atop shell 34, thereby closing opening 101 and prohibiting radiation from escaping upwards from cavity 26. Outlet air plenum 36 is formed between the bottom surface of lid 41 and the top of canister 70. Lid 41 comprises outlet ventilation ducts 42. Outlet ventilation ducts 42 form passageways from outlet air plenum 36, through lid 41, to the ambient air above ground level 212. Outlet ventilation ducts 42 do not have to be provided in lid 41, but can be formed as part of the integral structure 100 if desired. This will be discussed in greater detail below with respect to FIG. 14. Referring still to FIG. 13, when integral structure 100 is used to store spent nuclear fuel canister 70, the radiation shielding effect of the sub-grade is utilized while adequately facilitating cooling of canister 70. The cooling of canister 70 is facilitated by cool air entering inlet ventilation ducts 25 via above grade inlets 27. The cool air travels through inlet ventilation ducts 25 until it enters cavity 26 at or near inlet air plenum 33 via below grade outlets 28. Once the cool air is within cavity 26 it is warmed by the heat emanating from canister 70. As the air is warmed, it travels upward along the outer surface of canister 70 via annular space 60 until the air enters outlet air plenum 36. As the air travels upward through annular space 60 it continues to remove heat from canister 70. The warmed air then exits cavity 26 via outlet ventilation ducts 42 and enters the ambient air. This natural convective cooling flow repeats continuously until the canister 70 is adequately cooled. Referring now to FIG. 14, an alternative embodiment of an integral structure 200 is illustrated. Integral structure 200 is used to store a spent fuel canister in manner similar to that of integral structure 100 discussed above. While much of the structure is identical to that of integral structure 100, integral structure 200 further comprises outlet ventilation ducts 42 seal welded directly to shell 34. The outlet ventilation ducts 42 can be formed out of any of the materials discussed above with respect to the inlet ventilation ducts 25. As a result of the outlet ventilation ducts 42 being part of integral structure 200, lid 41 can be free of such ducts. The cooling process of canister 70 remains the same. FIG. 15 illustrates an integral structure 300 according to another aspect of the present invention. Integral structure 300 is similar in many respect to that of integral structures 100 and 200 in its design and functioning. However, integral structure 300 is specifically designed to store canisters 70 holding low heat spent fuel. When a canister 70 is giving off low heat, for example in the magnitude of 2-3 kW, it is not necessary to supply inlet ventilation ducts to supply cool air to cavity 26. Therefore, the inlet ventilation ducts are omitted from integral structure 300. Integral structure 300 comprises only outlet ventilation ducts 42, which act as both an inlet for the cooler air and an outlet for the warmer air. While outlet ventilation ducts 42 of integral structure 300 are seal welded to shell 34, it is possible for the outlet ventilation ducts to be located in the lid 41 if desired. Moreover, the concept of eliminating the inlet ventilation ducts for low heat load canister storage can be applied to any of the underground or above ground VVO embodiments illustrated in this application, specifically including underground VVO 20 and it derivatives. While the invention has been described and illustrated in sufficient detail that those skilled in this art can readily make and use it, various alternatives, modifications, and improvements should become readily apparent without departing from the spirit and scope of the invention. Specifically, it is possible for the entire underground VVO and/or integral structure of the present invention to be below grade, so long as the inlet ventilation ducts and/or outlet ventilation ducts open to the ambient air above grade. This facilitates very deep storage of spent fuel canisters. |
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041585992 | claims | 1. The method of refueling, in a relatively short time interval, a nuclear reactor having a pressure vessel, a core therein having fuel assemblies to be replaced; fuel in said fuel assemblies having an enrichment of about 2.7% to 3.2% in fissionable material, an upper package including the head of the pressure vessel, control rods and control rod drive mechanisms and housings, electrical cables connected at one end to said control rod drive mechanisms and adapted for connection at their other ends to a power source, a lifting device and said head being held to said vessel by studs having stud tensioners and detensioning means for detensioning said studs, the said method comprising the steps of: increasing the concentration of neutron absorbing material to its refueling magnitude in coolant circulated through the core, said refueling magnitude being that amount and concentration of neutron absorbing material which corresponds to the fissionable material enrichment in the fuel assemblies and which is needed to compensate for neutron activity upon withdrawal of control rods from the core; actuating said detensioning means to unlock and remove said studs thereby unlocking said head from said vessel, lifting said upper package in a single lifting operation to a prepared position away from said pressure vessel thus affording access to said core, flooding borated water equal to said refueling magnitude into a pit in which the reactor is located before said upper package is removed from the pressure vessel; raising the level of said water in the pit to a point just below the stud detensioners and cables, and then maintaining said level by uniformly increasing the level of said water in the pit during the time of lifting said upper package from the pressure vessel; and by virtue of such access replacing said fuel assemblies in said core at relatively short intervals by replacing about ten percent of the spent fuel assemblies when the reactor is refueled at three month intervals; and replacing about twenty percent of the spent fuel assemblies when the reactor is refueled at six month intervals. 2. The method according to claim 1 including the step of removing fuel assemblies from the center of the core and replacing them with fuel assemblies outside of the center, and replacing the assemblies from outside of the center with fresh assemblies, when the reactor is refueled at six month intervals; and replacing each removed assembly directly with a fresh assembly, without rearranging the fuel assemblies, when the reactor is refueled on a three month basis. |
abstract | The invention discloses a scanning method and apparatus suitable for scanning a pipeline or a process vessel in which a beam of gamma radiation from a source is emitted through the pipeline or the process vessel to be detected by an array of detectors, which are each collimated to detect gamma radiation over a narrow angle relative to a width of the emitted beam of gamma radiation. |
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047117591 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT FIG. 1 is an elevational view, in cross-section, of an ice basket 10 having a cylindrical, perforated metal sidewall and divided into a series of compartments, delineated by removable cruciforms 14 in accordance with the present invention. The open, upper end 12a of ice basket 10 affords limited access to the interior of the basket 10, the lower end 12b typically being enclosed by a grating or meshlike end closure (not shown) which is contiguous with support structure (not shown) for the basket 10. In a typical installation, the removable cruciforms 14 of the invention are disposed at axially displaced positions, or elevations, within the basket 10, corresponding to those of original, welded-in-place cruciforms, and thus are disposed at approximately 6 foot intervals, defining a succession of seven compartments 11-1 through 11-7 delineated by the plurality of cruciforms 14 within the basket 10, each containing initially a full charge of ice. As described in further detail hereinafter, each of the cruciforms 14 is releasably engaged on a stiffening ring (not seen in FIG. 1) for retaining same in position at the desired elevation within the ice basket 10. The cruciforms 14 are seen to perform the intended function of supporting the charges of ice within the corresponding compartments, despite the fact that sublimation has resulted in reduced charges of ice existing within the lower compartments. For example, whereas the charges of ice 9-1 and 9-2 substantially fill the corresponding compartments 11-1 and 11-2, in the lowermost compartments 11-6 and 11-7, significantly depleted charges of ice 9-6 and 9-7, respectively, remain. Whereas removal of the charge of ice at the uppermost compartment 11-1 is feasible with conventional tools since accessible through the open upper end 12a, removal of ice from the successively lower compartments is a difficult task. A preferred instrument which enables efficient and effective removal of ice from each of the successive compartments throughout the entire height of the basket 10 is disclosed in the copending application entitled "Ice Remover Auger for Ice Condention Containment", the inventors of which are coinventors herein, the application furthermore being assigned to the common assignee hereof. It thus will be understood that as the ice in each successive, lower compartment is removed, access may be gained to the corresponding removable cruciform 14 of the invention, and the same may be retracted and withdrawn, compartment by compartment, thereby to gain access to the lowermost compartment 11-7. The ice basket 10 may then be recharged with ice and the cruciforms 14 reinstalled, in compartment by compartment order. The cruciforms 14 of the invention, however, also accomodate alternative techniques and related equipment for accomplishing these same purposes, as later described herein. It furthermore should be understood that, where possible, removable cruciforms 14 in accordance with the present invention may be employed initially in a new installation, and not merely as a replacement for the conventional welded-in-place cruciforms of prior ice baskets, subsequently to their removal. However, where the removable cruciforms are to be employed in existing ice baskets as a replacement for the welded-in-place, conventional cruciforms, the latter must first be removed. Equipment for performing that function is disclosed in the copending application entitled "Ice Basket Cruciform Removal Tool", the inventor of which is a coinventor herein, the application being assigned to the common assignee hereof. The cruciform 14 of the present invention is shown in detail in the perspective, elevational view of FIG. 2 and, as assembled within an ice basket 10, in the plan view of FIG. 3, taken generally along the cross-sectional view line 3--3 in FIG. 1. Further details of the cruciform 14 of the invention are set forth in the side elevational view of FIG. 4, partly in cross-section and taken generally along the line 4--4 in FIG. 3, but wherein the sidewall 12 of the ice basket has been removed for simplicity and clarity of illustration. Additionally, the elevational view of FIG. 5 comprising a partial cross-section, taken along line 5--5 in FIG. 3, illustrates details of the internal construction of the cruciform 14. With concurrent reference to FIGS. 2, 3 and 4, the removable cruciform 14 in accordance with the present invention comprises a pair of brackets 16, each of a generally V-shaped, truncated base configuration. Each bracket 16 comprises a central, base portion 17 having parallel longitudinal edges 17-1 and 17-2 from which corresponding integral legs 18 extend at a predetermined angle, so as to assume generally radial orientations relative to the sidewall 12 of an ice basket 10 in which the cruciform 14 is installed, as best seen in FIG. 3. Each of the legs 18 carries a pair of integral, upper and lower feet 19 which extend radially beyond the outer longitudinal edge of the corresponding leg 18 and define a receiving channel 20 therebetween. A pair of supoort plate assemblies 22 and 24, respectively comprising parallel, spaced plates 22a, 22b, and 24a, 24b, defining corresponding slide channels 23 and 25 therebetween, are secured to and extend in parallel relationship from one of the brackets 16, at right angles to the central portion 17 thereof. As best seen in FIG. 4, a pair of parallel, horizontal slots 26 and 28 are formed so as to extend, in alignment, through each of the parallel support plates 22a, 22b and 24a, 24b. On the other of the brackets 16 there are secured a pair of slide support plates 32 and 34, affixed thereto so as to extend at right angles from the central portion 17 in parallel relationship, and spaced apart so as to be received in telescoping, sliding relationship in the corresponding slide channels 23 and 25 of the support plate assemblies 22 and 24. Each of the slide support plates 32 and 34 has secured thereto corresponding pin pairs 36 and 38, at positions aligned with, and for being received through, the slots 26 and 28, respectively, in the mating support plate assemblies 22 and 24, to restrict, or limit, the telescoping, sliding relationship to a direction parallel to the slots 26 and 28, and a length of travel as defined by the abutment of the pin pairs 36 and 38 with the corresponding, opposite ends of the slots 26 and 28. As can be best appreciated from FIG. 3, the pair of brackets 16 accordingly may be compressed and/or expanded with a limited length of travel along a diameter of the ice basket 10 passing perpendicularly through the respective central portions 17 thereof, and corresponding to a symmetrically disposed, compression/expansion axis of the cruciform 14. As best seen in FIG. 4, the slide support plates 32 and 34 are slightly shorter in axial height (i.e., along the vertical axis of the cruciform 14, corresponding to the vertical axis of the cylindrical basket 10) than the corresponding support plates 22a, 22b and 24a, 24b. For example, support plates 22a, 22b and 24a, 24b, may each be approximately 4 inches in axial height whereas the slide support plates 32 and 34 may be of approximately 3.62 inches in axial height. The V-shaped brackets 16 as well may be of approximately 4 inches in axial height. The central portions 17 of the respective brackets 16 and the telescopingly engaged support plate assemblies and slide support plates 22, 32, and 24, 34, define therewithin a spring housing 40 which is of nominally square cross-section but, as described, may be compressed or expanded within a limited extent of travel along an axis perpendicular to the parallel, central portions 17. A C-shaped spring 42 is received in the housing 40. With concurrent reference to FIGS. 3 to 7, the spring 42 defines a longitudinal, or axially extending opening 43 between its free ends 42a and 42b; while illustrated in FIG. 7 as of circular cross-sectional configuration, corresponding to its installed condition in FIG. 3, the spring 42 in a free configuration (i.e., when not disposed within the housing 40) assumes a normal, expanded configuration. Accordingly, the spring 42 engages the base portions of the brackets 16 and applies a resilient biasing force thereto for maintaining the spaced relationship thereof and the nominally square cross-section configuration of the housing 40, the pins 36 and 38 abutting the ends of the slots 26 and 28. Pairs of notches 44 and 46 are formed in the spring 42 at its opposite ends, each pair preferably being symmetrically disposed about a diameter which, as seen in FIG. 7, extends transversely to a diameter passing symmetrically through the opening 43. Pairs of tabs 48 and 50, as seen in FIGS. 3 and 5, are formed on the upper and lower edges of the central portions 17 of both brackets 16 and extend laterally and thus radially inwardly, so as to be received in the corresponding, aligned notches 44 and 46 of the spring 42. The cylindrical sidewall 12 of the ice basket 10 has rigidly secured therein, at spaced elevations, a plurality of stiffening rings 11. A single such stiffening ring 11 is illustrated in FIGS. 3 and 4, which may be welded in place and/or secured to the sidewall 12 by screws 13. In use of the replaceable cruciform 14 of the invention, the brackets 16 are moved together by compressing the spring 42, as before described, thus effectively retracting the radially extending legs 18. In a specific embodiment of the cruciform, the retraction or compression reduces the effective diameter of the circumferential periphery of the legs and associated feet extensions from 11.90" (as installed and engaged within the basket 10) to 11.50". This permits lowering the cruciform 14, in a horizontal orientation, axially downwardly through the ice basket 10 to a desired elevation in alignment with a stiffening ring 11. The cruciform 14 then is released from compression while supported at the desired elevation, the C-spring 42 causing the cruciform 14 to expand, advancing the legs 18 toward the sidewall 12 of the ice basket 10 so as to receive the stiffening ring 11 in the channels 20 intermediate the feet 19 of each leg 18. Suitable apparatus for effecting the compression of the bracket 14 and its axial, elevational positioning for installation in an engaged position with a stiffening ring 11, as well as for subsequent removal thereof, as described in the foregoing and illustrated in FIGS. 3 and 4, is disclosed in the concurrently filed and copending application entitled "Handling Tool for, and Method of Use of Ice Basket Removable Cruciform" of the common coinventors herewith and assigned to the common assignee hereof. The removable cruciform 14 of the oresent invention thus satisfies the requirement of being readily manipulated, both for installation into and removal from required elevations within an ice basket, for the purposes hereinbefore set forth. The configuration of the cruciform 14, moreover, is particularly advantageous, taking into account the maintenance functions required to be performed with respect to ice baskets of the type herein considered. Particularly, the cruciform 14 affords equivalent ice charge support functions, as those of the fixed, or welded-in-place, cruciforms of the prior art and, in fact, improves the support function in view of the generally square configuration of the spring housing 40, as compared to the relatively more simple, X-shaped configuration of the metal straps of the prior art cruciforms. Significantly, moreover, the mating configuration of the C-shape spring 42, as disposed within the housing 40, affords a central, axially aligned passage throughout the height of a given ice basket 10. Specifically, the spring 42, of approximately 3 inches in diameter for the embodiment as illustrated, when used with an ice basket of approximately one (1) foot diameter, affords a convenient central passageway or column, passing through the geometric center of the ice basket 10 throughout its height, to permit thermal drilling operations to provide an axially extending, central hole throughout the height of the ice basket through which maintenance tools may be inserted to remove and settle ice. The removable cruciform of the present invention thus will be seen to be highly effective and practical, not only as to its installation and removal and in performing its intended function as installed but also in facilitating certain maintenance operations and accommodating equipment employed for that purpose, yet is of relatively simple construction affording the benefits of ease and low cost of manufacture. Numerous modifications and adaptations of the removable cruciform insert of the present invention will be apparent to those of skill in the art and thus it is intended by the appended claims to cover all such modifications and adaptations which fall within the true spirit and scope of the invention. |
summary | ||
description | 1. Technical Field The present invention relates generally to semiconductor manufacturing and related technologies. More particularly, the present invention relates to electron beam lithography. 2. Description of the Background Art As is well-understood in the art, a lithographic process includes the patterned exposure of a resist so that portions of the resist can be selectively removed to expose underlying areas for selective processing such as by etching, material deposition, implantation and the like. Traditional lithographic processes utilize electromagnetic energy in the form of ultraviolet light for selective exposure of the resist. As an alternative to electromagnetic energy (including x-rays), charged particle beams have been used for high resolution lithographic resist exposure. In particular, electron beams have been used since the low mass of electrons allows relatively accurate control of an electron beam at relatively low power and relatively high speed. Electron beam lithographic systems may be categorized as electron-beam direct write (EBDW) lithography systems and electron beam projection lithography systems. In EBDW lithography, the substrate is sequentially exposed by means of a focused electron beam, wherein the beam either scans in the form of lines over the whole specimen and the desired structure is written on the object by corresponding blanking of the beam, or, as in a vector scan method, the focused electron beam is guided over the regions to be exposed. The beam spot may be shaped by a diaphragm. EBDW is distinguished by high flexibility, since the circuit geometries are stored in the computer and can be optionally varied. Furthermore, very high resolutions can be attained by electron beam writing, since electron foci with small diameters may be attained with electron-optical imaging systems. However, it is disadvantageous that the process is very time-consuming, due to the sequential, point-wise writing. EBDW is therefore at present mainly used for the production of the masks required in projection lithography. In electron beam projection lithography, analogously to optical lithography, a larger portion of a mask is illuminated simultaneously and is imaged on a reduced scale on a wafer by means of projection optics. Since a whole field is imaged simultaneously in electron beam projection lithography, the attainable throughputs can be markedly higher in comparison with electron beam writers. Disadvantages of conventional electron beam projection lithography systems includes that a corresponding mask is necessary for each structure to be exposed. The preparation of customer-specific circuits in small numbers is not economic, because of the high costs associated with mask production. One embodiment relates to a dynamic pattern generator for reflection electron beam lithography which includes conductive pixel pads, an insulative border surrounding each conductive pixel pad so as to electrically isolate the conductive pixel pads from each other, and conductive elements coupled to the conductive pixel pads for controllably applying voltages to the conductive pixel pads. The conductive pixel pads are advantageously cup shaped with a bottom portion, a sidewall portion, and an open cavity. Another embodiment relates to a pattern generating apparatus which includes a well structure with sidewalls and a cavity configured above each conductive pixel pad. The sidewalls may include alternating layers of conductive and insulative materials. Other embodiments, aspects and feature are also disclosed. Maskless Reflection Electron Beam Lithography As discussed above, electron-beam direct write (EBDW) lithography has the potential to achieve excellent resolution. However, EBDW has a traditional problem relating to its low throughput. For example, it may take ten to one hundred hours to inscribe an entire wafer using EBDW lithography. One previous approach to attempt to increase the throughput is by increasing the beam current. However, when the current density exceeds a certain threshold, electron-electron interactions cause the beam to blur. This patent application relates to a system and method of electron beam lithography that overcomes the above-discussed disadvantages and problems. Rather than focusing the electron beam into a tiny spot, the approach described herein floods the wafer with the electron beam. This enables use of a high beam current while keeping the beam current density at a level consistent with minimal electron-electron interactions. For example, an area roughly 0.1 millimeters (mm) wide may be illuminated. That area is several orders of magnitude larger than a traditional EBDW system that focuses the beam into a much smaller spot, for example, with a spot size on the order of tens of nanometers (nm) wide. A flood beam 0.1 mm wide would normally not provide a writing resolution sufficiently high for practical use in integrated circuit manufacturing. However, the system and method disclosed herein enables high-resolution writing by partitioning the flood beam into a multitude (for example, four million) of independently controllable beams. While others have tried building multiple columns with multiple sources to achieve multiple beams, there are various difficulties in that approach, including the difficulty of making the multiple columns behave uniformly. The system and method disclosed herein may be implemented using a single column and a single source. A conventional multi-beam system would require a large array of blankers to achieve a multitude of controllable beams from a single column, each blanker being a small and independently controllable element that can be switched on and off rapidly. However, it is quite problematic to build and control such a large array. For example, a blanker array for a conventional multi-beam system is not normally buildable using integrated circuits because such integrated circuits are opaque to electrons. The system and method disclosed herein re-directs the beam out of the direct line of sight between the electron source and the semiconductor wafer. Independently-controllable voltages are applied to cells of a dynamic pattern generator array that may be implemented using integrated circuit technology. The voltages determine whether each cell reflects electrons onto the wafer or absorbs electrons (preventing them from being reflected onto the wafer). The system and method disclosed herein advantageously breaks through the traditional EBDW speed-versus-resolution tradeoff by illuminating a large area and simultaneously exposing a multitude of pixels on the wafer. For example, four million pixels may be exposed using a 4000×1000 array of individually addressable elements. This may be achieved using a single column and a conventional electron source. FIG. 1 is a schematic diagram of a maskless reflection electron beam projection lithography system 100 in accordance with an embodiment of the invention. The name may be abbreviated to a reflection electron beam lithography or REBL system. As depicted, the system 100 includes an electron source 102, illumination electron-optics 104, a magnetic prism 106, an objective electron lens 110, a dynamic pattern generator (DPG) 112, projection electron-optics 114, and a stage 116 for holding a wafer or other target to be lithographically patterned. In accordance with an embodiment of the invention, the various components of the system 100 may be implemented as follows. The electron source 102 may be implemented so as to supply a large current at low brightness (current per unit area per solid angle) over a large area. The large current is to achieve a high throughput rate. Preferably, the material of the source 102 will be capable of providing a brightness of about 104 or 105 A/cm2 sr (Amperes per cm2 steradian). One implementation uses LaB6, a conventional electron emitter, which typically have a brightness capability of about 106 A/cm2 sr, as the source material. Another implementation uses tungsten dispenser emitters, which typically have a brightness capability of about 105 A/cm2 sr when operating at 50 kilovolts, as the source material. Other possible emitter implementations include a tungsten Schottky cathode, or heated refractory metal disks (i.e. Ta). The electron source 102 may be further implemented so as to have a low energy spread. The REBL system 100 should preferably control the energy of the electrons so that their turning points (the distance above the DPG 112 at which they reflect) are relatively constant, for example, to within about 100 nanometers. To keep the turning points to within about 100 nanometers, the electron source 102 would preferably have an energy spread no greater than 0.5 electron volts (eV). LaB6 emitters have typical energy spreads of 1 to 2 eV, and tungsten dispenser emitters have typical energy spreads of 0.2-0.5 eV. In accordance with one embodiment of the invention, the source 102 comprises a LaB6 source or tungsten Schottky emitter that is operated a few hundred degrees C below its normal operating temperature to reduce the energy spread of the emitted electrons. However, cooler operating temperatures can destabilize the source 102, for example, due to impurities settling on the source surface and thereby diminishing its reliabilty and stability. Therefore, the source material may be preferably selected to be a material in which impurities are unlikely to migrate to the surface and choke off emission. Moreover, the vacuum on the system may be made stronger to overcome the impurity problem. Conventional lithography systems operate at a vacuum of 10−6 Torr. A scanning electron microscope (SEM) with a LaB6 source typically operates at 10−7 Torr. A SEM with a Schottky emitter typically operates at 10−9 Torr or better in the gun region. In accordance with one implementation, the REBL operates with a gun region vacuum of 10−9 Torr or lower to protect the stability of the source 102. The illumination electron-optics 104 is configured to receive and collimate the electron beam from the source 102. The illumination optics 104 allows the setting of the current illuminating the pattern generator structure 112 and therefore determines the electron dose used to expose the substrate. The illumination optics 104 may comprise an arrangement of magnetic and/or electrostatic lenses configured to focus the electrons from the source 102 so as to generate an incident electron beam 105. The specific details of the arrangement of lenses depend on specific parameters of the apparatus and may be determined by one of skill in the pertinent art. The magnetic prism 106 is configured to receive the incident beam 105 from the illumination optics 104. When the incident beam traverses the magnetic fields of the prism, a force proportional to the magnetic field strengths acts on the electrons in a direction perpendicular to their trajectory (i.e. perpendicular to their velocity vectors). In particular, the trajectory of the incident beam 105 is bent towards the objective lens 110 and the dynamic pattern generator 112. In a preferred embodiment, the magnetic prism 106 is configured with a non-uniform magnetic field so as to provide stigmatic focusing, for example, as disclosed in U.S. patent application Ser. No. 10/775,646, entitled “Improved Prism Array for Electron Beam Inspection and Defect Review,” filed Feb. 10, 2004 by inventor Marian Mankos, the disclosure of which is hereby incorporated by reference in its entirety. A uniform magnetic field provides astigmatic focusing wherein focusing occurs in only one direction (for example, so as to image a point as a line). In contrast, the magnetic prism 106 configuration should focus in both directions (so as to image a point as a point) because the prism 106 is also utilized for imaging. The stigmatic focusing of the prism 106 may be implemented by dividing it into smaller sub-regions with different but uniform magnetic fields. Furthermore, the lens elements in the prism 106 may be of a relatively longer length and width so as to provide for a low distortion image over a large field size. However, increasing the length of the prism 106 involves a trade-off of more electron-electron interactions causing more blur. Hence, the reduced image distortion should be balanced against the increased blur when increasing the prism length. Below the magnetic prism 106, the electron-optical components of the objective optics are common to the illumination and projection subsystems. The objective optics may be configured to include the objective lens 110 and one or more transfer lenses (not shown). The objective optics receives the incident beam from the prism 106 and decelerates and focuses the incident electrons as they approach the DPG 112. The objective optics is preferably configured (in cooperation with the gun 102, illumination optics 104, and prism 106) as an immersion cathode lens and is utilized to deliver an effectively uniform current density (i.e. a relatively homogeneous flood beam) over a large area in a plane above the surface of the DPG 112. In one specific implementation, the objective lens 110 may be implemented to operate with a system operating voltage of 50 kilovolts. Other operating voltages may be used in other implementations. The dynamic pattern generator 112 comprises an array of pixels. Each pixel may comprise a metal contact to which a voltage level is controllably applied. The principle of operation of the DPG 112 is described further below in relation to FIGS. 3A and 3B. The extraction part the of the objective lens 110 provides an extraction field in front of the DPG 112. As the reflected electrons 113 leave the DPG 112, the objective optics is configured to accelerate the reflected electrons 113 toward their second pass through the prism 106. The prism 106 is configured to receive the reflected electrons 113 from the transfer lens 108 and to bend the trajectories of the reflected electrons towards the projection optics 114. The projection electron-optics 114 reside between the prism 106 and the wafer stage 116. The projection optics 114 is configured to focus the electron beam and demagnify the beam onto photoresist on a wafer or onto another target. The demagnification may range, for example, from 1× to 20× demagnification (i.e. 1× to 0.05× magnification). The blur and distortion due to the projection optics 114 is preferably a fraction of the pixel size. In one implementation, the pixel size may be, for example, 22.5 nanometers (nm). In such a case, the projection optics 114 preferably has aberrations and distortions of less than 10-20 nm. The wafer stage 116 holds the target wafer. In one embodiment, the stage 116 is stationary during the lithographic projection. In another embodiment, the stage 116 is in linear motion during the lithographic projection. In the case where the stage 116 is moving, the pattern on the DPG 112 may be dynamically adjusted to compensate for the motion such that the projected pattern moves in correspondence with the wafer movement. In other embodiments, the REBL system 100 may be applied to other targets besides semiconductor wafers. For example, the system 100 may be applied to reticles. The reticle manufacturing process is similar to the process by which a single integrated circuit layer is manufactured. FIG. 2 is a schematic diagram of a maskless reflection electron beam projection lithography system 200 showing further components in accordance with an embodiment of the invention. The additional components illustrated include a high voltage source 202, a parallel datapath 204, an interferometer 206, a height sensor 208, feedback circuitry 210, and beam deflectors 212. The high voltage source 202 is shown as providing a high voltage to the source 102 and to the DPG 112. The voltage provided may be, for example, 50 kilovolts. The parallel data path 204 is configured to carry control signals to the DPG 112 for controlling the voltage on each pixel (so that it either absorbs electrons or reflects them). In one embodiment, the control signals are adjusted so that the pattern moves electronically across the DPG pixel array in a manner that is substantially the same as the way signals move through a shift register and at a rate so as to match the linear movement of the wafer. In this embodiment, each exposed point on the wafer may receive reflected electrons from an entire column (or row) of DPG pixels, integrated over time. In one implementation of this embodiment, the DPG 112 is configured to resemble a static random access memory (SRAM) circuit, such as that depicted in FIG. 4. In another embodiment, the control signals are such that the DPG 112 exposes one complete frame at a time. In this embodiment, each pixel on the DPG 112 exposes a corresponding pixel on the wafer. The pattern on the DPG 112 remains constant during the exposure of each frame. In one implementation of this embodiment, the DPG 112 is configured to resemble a dynamic random access memory (DRAM) circuit, such as that depicted in FIG. 5. The interferometer 206 may be included to provide tight coupling and positional feedback between the electron beam location and the target on the wafer. In one embodiment, the optical beams are reflected off mirrors on the stage. The resulting interference pattern depends on the difference of the individual beam paths and allows accurate measurement of the stage and wafer position. As further described below in relation to FIG. 6, optical beams from the interferometer are diffracted from a grating on the target substrate surface, the resultant interference pattern is sensitive to the phase of the grating and can therefore deliver lateral positional information. Vertical positional information may be provided by a height sensor 208. The positional information may be fed back via feedback circuitry 210 so as to control beam deflectors 212. The deflectors 212 are configured to deflect the projected beam so as to compensate for vibrations and positional drift of the wafer. FIGS. 3A and 3B are diagrams illustrating the operation of a dynamic pattern generator in accordance with an embodiment of the invention. FIG. 3A shows a cross-section of a DPG substrate 302 showing a column (or row) of pixels. Each pixel includes a conductive area 304. A controlled voltage level is applied to each pixel. In the example illustrated in FIG. 3A, four of the pixels are “off” and have zero (0) volts applied thereto, while one pixel (with conductive area labeled 304x) is “on” and has one (1) volt applied thereto. (The specific voltages may vary depending on the parameters of the system.) The resultant local electrostatic equipotential lines 306 are shown, with distortions 306x relating to “off” pixel shown. In this example, the incident electrons 308 approaching the DPG 112 come to a halt in front of and are reflected by each of the “on” pixels, but the incident electrons 308x are drawn into and absorbed by the “off” pixel. The resultant reflected current (in arbitrary units) is shown in FIG. 3B. As seen from FIG. 3B, the reflected current is “0” for the “off” pixel and “1” for the “on” pixels. FIG. 4 is a schematic diagram of a dynamic pattern generator 112 implemented in a circuit structure resembling a SRAM in accordance with an embodiment of the invention. In one embodiment, the SRAM-like DPG 112 is used in a “rolling mode” wherein the lithographic pattern is moved from one row of pixels into the next at the same rate at which the wafer scans. To expose a spot on the photoresist on a wafer, each pixel turns “on” (i.e. become reflective) as the spot passes “beneath” the pixel. FIG. 5 is a schematic diagram of a dynamic pattern generator 112 implemented in a circuit structure resembling a DRAM in accordance with an embodiment of the invention. Here, each pixel may be implemented with as few as one transistor and one capacitor, so smaller pixels may be implemented. In one embodiment, the DRAM-like DPG 112 is used in “frame” (or “step and flash”) mode. In frame mode, the DPG 112 exposes one frame at a time. Each DPG pixel exposes a corresponding pixel on the wafer, and the pattern on the DPG 112 remains constant during the exposure of each frame. As described above, the DPG 112 comprises a programmable pattern generator structure. In an alternate embodiment, the system 100 may utilize a static patterned structure. As described above, the DPG 112 is utilized to generate a pattern with pixels either on or off. In an alternate embodiment, gray scale values for the pixels may be implemented. Gray scale values may be implemented, for example, by using a finer range of voltage values so as to provide control over the percentage (between 0% and 100%) of electrons reflected by a pixel. Alternatively, gray scale values may be implemented by varying the percentage of time over which each pixel remains on. FIG. 6 is a schematic diagram depicting a feedback loop system for controlling electron beam position in accordance with an embodiment of the invention. Here, the electron beam is used to write a pattern 602 in a center portion of a reticle plate or wafer. As shown, the feedback loop system may be implemented using a low-profile waveguided interferometer attached to the projection lens. Dual incident laser beams may be diffracted from a grating pattern 604 into a common first-order axial beam. The resultant interference pattern is sensitive to the phase of the grating and can therefore deliver lateral (both x and y) positional information with sub-nanometer resolution. The grating pattern 604 may be, for example, placed around the edge of a reticle or in the scribe lines on a wafer. The grating pattern 604 may be implemented as a latent image on the photoresist, or as a feature inscribed on the periphery of a bare wafer or reticle blank. The system feeds back positional information to deflectors that steer the electron beam so as to cancel vibrations and positional drift. In accordance with another embodiment of the invention, a flat or ribbon electron beam is generated by the illumination electron optics. This ribbon type beam may have an aspect ratio of, for example, 1000:1 or 2000:1 so as to be substantially one dimensional. The reflective array may be a one-dimensional array. Advantageously, the one-dimensional reflective array and clocking signals for it may be simpler to implement. Low-Energy Backscattering Limitation and Solution Prior DPG designs had a flat structure, such as that depicted in FIGS. 7 and 8. As discussed below, this flat structure has a limitation due to the backscattering of low energy electrons. FIG. 7 is an electron micrograph of a surface of a flat dynamic pattern generator structure. The micrograph shows a planar assembly of conductive flat “pixel” pads 702 separated by insulating material borders 704. As shown, in one implementation, the surface area of each pixel may be very roughly 1 micron by 1 micron, and the insulating border may be roughly 0.1 microns wide. FIG. 8 is a schematic diagram showing a cross-sectional view of a flat dynamic pattern generator structure. As seen, each pixel pad 702 is formed as a solid conductive plug, and the insulative borders 704 are configured to electrically separate the pixels. The insulative borders 704 may be trenches which are partially filled with dielectric material. Voltage may be applied individually via conductive elements 802 to each pixel pad 702. The voltage applied to each flat pixel pad 702 may be controllably varied so as to either repel the incident electrons or to let them reach the surface and be absorbed. Landing energies for operation may be slightly above the higher end of the electron beam thermal distribution, typically at 3 to 5 electron volts (eV). However, applicants have determined disadvantageous aspects of flat-structured pixel pads 702. For example, the absorbing performance of a flat pixel pad 702 may be substantially affected by back scattering of low-energy electrons (for example, 3 to 5 eV in energy) from the pixel pad's surface. Such backscattering results in the generation of back-scattered (and hence detected) electrons in a regime where no reflected electron flux is expected from the corresponding pixel. This generates noise in the detected data and is an impediment in providing high contrast between the reflective and absorptive pixels of a DPG. For flat structured pixels, the backscattering of low electrons is found to be substantial and sensitive to the choice of the pixel pad material. In addition, strong variations are observed within a set of samples of a same pixel pad material. To overcome the above-discussed low-energy electron backscattering problem, the present application discloses a dynamic pattern generator with a cup-shaped (well-shaped) structure for the pixel pads. In other words, the DPG comprises pixel pads shaped as Faraday micro-cups. Applicants have determined that such cup-shaped pixel pads are advantageous in suppressing the undesirable backscattering effects discussed above. Also disclosed is a method of manufacturing the cup-shaped pixel pads on the surface of a DPG. FIG. 9 is a schematic diagram a cross-sectional view of a dynamic pattern generator structure with cup-shaped pixel pads in accordance with an embodiment of the invention. As shown, each pixel pad 902 in this structure is cup-shaped with a bottom portion and side walls so as to form a “cup” with an open volume or cavity 904. Insulative borders 906 are configured to electrically separate the cup-shaped pixel pads 902. Voltages may be applied via conductive elements 908 to the pixel pads 902. The inner profile geometry for each Faraday cup-shaped pixel pad 902 may vary from conical to cylindrical, depending on performance parameters. The conductive material to form the cup-shaped pixel pads may be a highly conductive material such as, for example, titanium (Ti), titanium nitride, tungsten (W), tantalum (Ta), aluminum (Al), or copper (Cu). FIGS. 10A and 10B are schematic diagrams illustrating steps in a process for fabricating a cup-shaped pixel pad for a dynamic pattern generator structure in accordance with an embodiment of the invention. While specific process steps are shown in FIGS. 10A and 10B, additional or alternative process steps may be used in other implementations. First, on the left in FIG. 10A, a diagram shows a cross-sectional view of the structure being manufactured after an etch step which forms a well 1002 for each pixel in the insulative layer but leaves an insulative border 906 around each well to electrically separate the pixel pads. A previously-formed bottom layer with the conductive element 908 used to controllably apply a voltage to the pixel pad is also shown. Second, in the middle of FIG. 1A, a cross-sectional diagram shows the structure after deposition of the conductive material for the cup-shaped pixel pad 902. For example, a metallic physical vapor deposition process may be used to deposit the conductive material for the pixel pad 902 such that the pixel pad 902 has a “cup-shaped” opening or cavity 1004, rather than being flat. Third, on the right in FIG. 10A, a cross-sectional diagram shows the structure after a fill deposition which fills 1006 the opening 1004. The material used to fill the opening is such that it may be etched differentially from the conductive material of the pixel pad 902 in a later step. Fourth, on the left in FIG. 10B, a cross-sectional diagram shows the structure after chemical-mechanical planarization (CMP) of the surface. This results in a flat or planarized surface 1008. Fifth, on the right in FIG. 10B, a cross-sectional diagram shows the structure after an etch step in which the fill 1006 and the dielectric 906 are selectively etched, while the cup-shaped conductive material of the pixel pad 902 is not etched. This results in the same structure as shown in FIG. 9. The cup-shaped pixel pad design described above has various advantages. First, it reduces or minimizes the DPG performance sensitivity to pixel pad material choice. Second, it reduces the requirements relating to pixel pad surface contamination and its possible unintended consequence of chemical modification. As a result, the requirements relating to handling of the DPG and vacuum conditions in the DPG chamber may be relaxed. Third, the inner cavity shape of the Faraday cup may be adjusted so as to adjust or optimize the shape of the near-DPG electrical field and so adjust or optimize the DPG performance. Microlensing Limitation and Solution Another limitation relates to an effect called microlensing or interlensing which occurs because of the energy spread of the incident electron beam. As discussed above, for example, in relation to FIGS. 3A and 3B, each of the pads may have a voltage applied so that it either absorbs or reflects incident electrons. Applicants have determined that the prior DPG design with flat-shaped pixel pads 702 as shown in FIG. 7 experiences a microlensing or interlensing effect in which electrons to be reflected by the pixel pads at negative potential may behave differently depending upon the potentials at the neighboring pads. The microlensing or interlensing effect affects the linearity of the DPG, making it difficult to implement a gray-level writing implementation. One potential strategy for mitigating the microlensing effect involves a high immersion electrostatic field applied to the surface of the DPG 112 by the objective electron optics 110 combined with a low energy spread of the incident electron beam 105 (see FIG. 1). By increasing the value of the applied electrostatic field, the electrostatic cross talk between the pixel pads decreases, and by decreasing the energy spread of the incident electron beam, most of the reflected electrons should be reflected near the surface of the pad, potentially minimizing the microlensing. However, the above potential strategy has some practical limitations. First, an electrostatic field that exceeds a certain value (for example, 200 volts per millimeters) is difficult to achieve without compromising (reducing) the field of view (FOV) of the illuminating (incident) electron beam. A higher electrostatic field will generally introduce more distortions affecting the FOV size and thus the system throughput would be reduced. In addition, increasing the electrostatic field applied by the objective lens also decreases the depth of field (DOF) of the illuminating electron beam. Moreover, a lower energy spread generally requires a lower beam current, also reducing the throughput. The microlensing effect and its non-linearity may be shown by comparing a simultaneous “aerial” image formed by four pixels in a 2×2 arrangement simultaneously “on” (i.e. in a reflecting state) with a summation “aerial” image of formed by summing four separate aerial images, each component image being obtained by turning “on” one pixel at a time. For example, FIG. 11 is a graph showing a comparison between the peak scatter factor (PSF) of four pixels simultaneously on and the PSF of the four pixels summed for a flat pixel pad design. As seen in FIG. 11, the “simultaneous” case had a significantly higher sigma of 180 nanometers (nm) compared with the “summed” case which had a lower sigma of 130 nm. This shows evidence of the microlensing effect for the flat pixel pad design. FIG. 12 is a schematic diagram showing a cross-sectional view of a dynamic pattern generator having a cup-shaped (well-shaped) pixel structure which includes stacked electrodes configured to collect, focus, and extract electrons in accordance with an embodiment of the invention. In this design, the sidewalls of each well (cup) is a stack with alternating insulative (for example, 1202, 1204, and 1206) and conductive (for example, 1212, 1214, and 1216) layers. In the specific embodiment depicted, the metal pad 1210 at the bottom of each well is about 1.4 microns across. The lower conductive layer 1212 (about 1 micron above and closest to the metal pad at the bottom of the well) has an applied voltage of negative 2.5 volts (V), the middle conductive layer 1214 (about 1 micron above the lower conductive layer) has an applied voltage of positive 15 V, and the upper conductive layer 1216 (about 1 micron above the middle conductive layer) has an applied voltage of positive 0.5 volts. For example, the reflecting/absorbing metal pad 1210 at the bottom of the well may have a voltage swing of 0 volts (for absorbing) to negative 4 volts (for reflecting). The upper conductive layer (electrode) 1216 may have a relatively weak positive applied voltage of positive 0.5 volts to both screen the insulator from the incoming electron current and to deflect the incoming electrons with lower energy towards the inside of a nearest well. The middle conductive layer (electrode) 1214 may have a relatively strong positive applied voltage of positive 15 volts so as to both focus the incoming electrons by drawing them into the well and extracting the reflected electrons by drawing them out of the well. The lower conductive layer (electrode) 1212 may have a negative applied voltage of negative 2.5 volts so as to focus the reflected electrons. Of course, specific implementations may use different configurations of stacked electrodes and/or different specific voltages to perform the same or similar functionalities. FIG. 13 is a graph showing how the stacked electrode design results in similar sigmas for the simulation with four pixels simultaneously on and the simulation with PSF of the four pixels summed for a well-shaped pixel pad design with stacked electrodes. As seen in FIG. 13, the “summed” and “simultaneous” cases had very similar sigmas (ranging from a low point of about 38 or 39 nm) when taking into account a minimal shift in the focus position at the object plane. This means that each well returned a reflected beam whose characteristics are independent or nearly independent from the state of the neighboring wells. Hence, the microlensing or interlensing effect is largely mitigated by the stacked electrode design of FIG. 12. FIGS. 14A and 14B depict electron trajectory simulations showing that a numerical aperture (NA) of the reflected beam may be adjusted by changing the voltage of the lower stacked electrode 1212 in accordance with an embodiment of the invention. The simulation depicted in FIG. 14A adjusted the applied voltage to negative 3.5 volts on the lower electrode 1212. The simulation depicted in FIG. 14B adjusted the applied voltage to negative 1.5 volts on the lower electrode 1212. Advantageously, such adjustments may be utilized to adjust the numerical aperture (NA) or pixel semiangle to potentially compensate for space charge induced aberrations. Advantageously, the above-described well-shaped pixel structure with stacked electrodes substantially mitigates microlensing effects, such that linear proximity effect correction algorithms may be used. The structure further works well with very low electrostatic fields and so avoid the need for a large immersion field, as well as avoiding the need to reduce the FOV and DOF. It can further work well with relatively large energy spread (for example, an energy spread of 3 eV at full width half maximum), and so enable the use of relatively large illumination currents. In addition, by changing one or more of the applied voltages on the stacked electrodes, space charge induced aberrations may be corrected. The above-described diagrams are not necessarily to scale and are intended be illustrative and not limiting to a particular implementation. In the above description, numerous specific details are given to provide a thorough understanding of embodiments of the invention. However, the above description of illustrated embodiments of the invention is not intended to be exhaustive or to limit the invention to the precise forms disclosed. One skilled in the relevant art will recognize that the invention can be practiced without one or more of the specific details, or with other methods, components, etc. In other instances, well-known structures or operations are not shown or described in detail to avoid obscuring aspects of the invention. While specific embodiments of, and examples for, the invention are described herein for illustrative purposes, various equivalent modifications are possible within the scope of the invention, as those skilled in the relevant art will recognize. These modifications can be made to the invention in light of the above detailed description. The terms used in the following claims should not be construed to limit the invention to the specific embodiments disclosed in the specification and the claims. Rather, the scope of the invention is to be determined by the following claims, which are to be construed in accordance with established doctrines of claim interpretation. |
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abstract | A scintillator array includes a first scintillator element, a second scintillator element, and a reflector provided between the first and second scintillator elements and having a width of 80 μm or less therebetween. Each scintillator element includes a polycrystal containing a rare earth oxysulfide phosphor, the polycrystal having a radiation incident surface of 1 mm or less×1 mm or less in area. An average crystal grain diameter of the polycrystal is not less than 5 μm nor more than 30 μm, the average crystal grain diameter being defined by an average intercept length of crystal grains in an observation image of the polycrystal with a scanning electron microscope. A maximum length or a maximum diameter of defects on the polycrystal is 40 μm or less. |
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053496175 | abstract | A pressured-water nuclear reactor comprising an apparatus for directly removing the residual power from the core of the reactor is provided. Advantageously, upon shut-down of the nuclear reactor of the present invention, residual power of the reactor may be safely removed therefrom. |
abstract | A source of photons or neutrons includes a housing that defines a discharge chamber, a first group of ion beam sources directed toward a plasma discharge region in the discharge chamber, the first group of ion beam sources including a first electrode and an inner shell, and a second electrode spaced from the plasma discharge region. The source of photons or neutrons further includes a first power supply for energizing the first group of ion beam sources to electrostatically accelerate toward the plasma discharge region ion beams which are at least partially neutralized before they enter the plasma discharge region, and a second power supply coupled between the first and second electrodes for delivering a heating current to the plasma discharge region. The ion beams and the heating current form a hot plasma that radiates photons or neutrons. The source of photons or neutrons may further include a second group of ion beam sources. The photons may be in the soft X-ray or extreme ultraviolet wavelength range and, in one embodiment, have wavelengths in a range of about 10-15 nanometers. |
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description | The present invention claims priority under 35 U.S.C. §119 to Japanese Application No. 2012-91312 filed Apr. 12, 2012, the entire content of which is incorporated herein by reference. 1. Technical Field The present invention relates to an X-ray detection apparatus for irradiating a sample with X-rays and detecting fluorescent X-rays generated from the sample. 2. Description of Related Art X-ray fluorescence analysis is an analytical method including steps of: irradiating a sample with X-rays; detecting fluorescent X-rays generated from the sample; and making a qualitative analysis or a quantitative analysis of elements contained in the sample according to a fluorescent X-ray spectrum. In general, an X-ray detection apparatus is provided with a collimator prepared by forming an aperture at an object which blocks X-rays in order to limit the range of X-rays by narrowing X-rays. An X-ray detection apparatus, which is equipped with a collimator including a plurality of apertures having different diameters and can move the collimator, has been developed so that the diameter of an aperture can be changed depending on the type of a sample, the purpose of analysis or the like. In an X-ray detection apparatus, X-rays other than fluorescent X-rays generated from a sample, such as scattered X-rays or fluorescent X-rays generated at a collimator, are sometimes detected. Since detection of such X-rays causes reduction in accuracy of X-ray fluorescence analysis, it is preferable to minimize the detection of X-rays other than fluorescent X-rays generated from a sample in order to realize highly accurate X-ray fluorescence analysis. A technique disclosed in Japanese Patent Application Laid-Open No. 2010-66121 is an X-ray diffractometer equipped with a shield for blocking extra X-rays. There is a need for downsizing of an X-ray detection apparatus in order to respond to the microminiaturization of a sample such as a circuit board. There is also a demand for downsizing of an X-ray detection apparatus in order to realize easier X-ray fluorescence analysis. In a downsized X-ray detection apparatus, an X-ray irradiation unit, an X-ray detector, a sample support unit and a collimator are located as proximally as practicable to each other. In such an X-ray detection apparatus, X-rays other than fluorescent X-rays generated from a sample, such as X-rays which are to enter the X-ray detector directly from the X-ray irradiation unit, tend to enter the X-ray detector more easily. Therefore, it is necessary to block X-rays other than fluorescent X-rays from a sample effectively. The present invention has been made in view of such problems, and the object thereof is to provide a downsized X-ray detection apparatus which can block X-rays other than fluorescent X-rays from a sample effectively by locating a shield appropriately. An X-ray detection apparatus according to the present invention is an X-ray detection apparatus comprising: a sample support unit; an X-ray irradiation unit configured to irradiate a sample supported by the sample support unit with X-rays; an X-ray detector configured to detect X-rays generated from the sample; and a collimator configured to narrow X-rays to be used for irradiation of the sample by the X-ray irradiation unit, characterized in that the X-ray irradiation unit and the X-ray detector are located with an exit of X-rays at the X-ray irradiation unit and an entrance of X-rays at the X-ray detector faced to a predetermined part of the sample support unit, and a shield, which blocks X-rays passing through a path linking the exit with the entrance and X-rays passing through a path linking an arbitrary part of the collimator with the entrance, is equipped. In the present invention, an X-ray detection apparatus for detecting X-rays from a sample is equipped with an X-ray irradiation unit, an X-ray detector and a shield for blocking X-rays. The shield blocks X-rays, which are to enter the X-ray detector directly from the X-ray irradiation unit, and fluorescent X-rays and scattered X-rays generated at the collimator. An X-ray detection apparatus according to the present invention is characterized in that the shield includes: a first shielding member configured to block X-rays passing through a path linking the exit with the entrance; and a second shielding member configured to block X-rays passing through a path linking the first shielding member and the collimator with the entrance. In the present invention, the shield includes: a first shielding member for blocking X-rays, which are to enter the X-ray detector directly from the X-ray irradiation unit; and a second shielding member for blocking X-rays from the collimator and the first shielding member so as to prevent the X-rays from entering the X-ray detector. An X-ray detection apparatus according to the present invention is characterized in that the collimator has a plate-like shape, and the shield is projected from both faces of the collimator. In the present invention, the shield is projected to both face sides of the plate-like collimator. In such a structure, X-rays from both faces of the collimator are also blocked by the shield. An X-ray detection apparatus according to the present invention is characterized in that the shield has a shape not to block an X-ray path from a sample supported by the sample support unit to the entrance. In the present invention, the shield does not block fluorescent X-rays, which are generated at the sample and are detected by the X-ray detector. Accordingly, fluorescent X-rays of a sample are detected efficiently. An X-ray detection apparatus according to the present invention is characterized in that the shield and the collimator are joined with each other. In the present invention, the shield is joined with the collimator. An X-ray detection apparatus according to the present invention is characterized in that the collimator includes a plurality of apertures configured to narrow X-rays, and the shield and the collimator can move to change an aperture through which X-rays pass. In the present invention, the collimator can move to change an aperture for narrowing X-rays, and the collimator and the shield move as a unit. With the present invention, a shield prevents X-rays other than fluorescent X-rays generated from a sample from entering an X-ray detector, and reduction in accuracy of X-ray fluorescence analysis is prevented, even in a state where an X-ray irradiation unit, the X-ray detector, a sample support unit and a collimator are located proximally to each other. Consequently, the present invention has beneficial effects such that it becomes possible to downsize an X-ray detection apparatus which can be used for highly accurate X-ray fluorescence analysis. The above and further objects and features of the invention will more fully be apparent from the following detailed description with accompanying drawings. The following description will explain the present invention concretely with reference to the drawings for illustrating an embodiment thereof. FIG. 1 is a schematic perspective view for illustrating the main structure of an X-ray detection apparatus. An X-ray detection apparatus is an apparatus for making an X-ray fluorescence analysis including steps of: detecting fluorescent X-rays generated by irradiating a sample with X-rays; and measuring a fluorescent X-ray spectrum or analyzing elements contained in the sample. An X-ray detection apparatus is equipped with a sample support unit 1 for supporting a sample. The sample support unit 1 has a horizontal plate shape and supports a sample as the sample is placed thereon. The sample support unit 1 is provided with a through hole 11. A sample, which is not illustrated in FIG. 1, is placed to close the through hole 11. FIG. 2 is a schematic sectional view for illustrating the II-II cross section of FIG. 1. A sample S is placed at a position to close the through hole 11 of the sample support unit 1. An X-ray irradiation unit 4 for irradiating the placed sample S with X-rays, a collimator 2 for narrowing X-rays from the X-ray irradiation unit 4, and an X-ray detector 5 for detecting fluorescent X-rays generated from the sample S are located below the sample support unit 1. Moreover, the collimator 2 is joined with a shield 3 for blocking X-rays. Although the X-ray irradiation unit 4 and the X-ray detector 5 are illustrated in FIG. 2 as cross sections simplistically, the X-ray irradiation unit 4 and the X-ray detector 5 are actually composed of a plurality of components and also include cavities therein. The sample support unit 1 has a base unit 13 and a detachable unit 12 which can be attached to and detached from the base unit 13. The through hole 11 is formed both at the base unit 13 and the detachable unit 12, and the base unit 13 and the detachable unit 12 form a substantially plate-like shape. An X-ray transparent film 14 is spread to close the through hole 11, and the X-ray transparent film 14 is fixed between the base unit 13 and the detachable unit 12. The X-ray transparent film 14 is fixed by steps of: spreading the X-ray transparent film 14 at the through hole 11 of the base unit 13 with the detachable unit 12 detached; and attaching the detachable unit 12 to the base unit 13. The sample S is placed above the X-ray transparent film 14. The X-ray irradiation unit 4 is located at a position to irradiate an undersurface of the sample S, which is placed on the sample support unit 1, with X-rays from obliquely below. The X-ray irradiation unit 4 is constituted of an X-ray tube and is located with an exit end of X-rays faced to the through hole 11 of the sample support unit 1. The X-ray detector 5 is located at a position to detect fluorescent X-rays radiated from the undersurface of the sample S, which is placed on the sample support unit 1, obliquely downward. The X-ray detector 5 is constituted of an X-ray detection element such as a silicon device and is located with an entrance end of fluorescent X-rays faced to the through hole 11 of the sample support unit 1. That is, the X-ray irradiation unit 4 and the X-ray detector 5 are located on the same face side of the plate-like sample support unit 1 and are located with the exit of X-rays and the entrance of fluorescent X-rays faced to a common predetermined part of the sample support unit 1. It is to be noted that there is no need to make the axis of irradiation of the X-ray irradiation unit 4 and the axis of entrance of the X-ray detector 5 overlap with each other at one point of the sample support unit 1 but it is only preferable to make the range of irradiation of X-rays from the X-ray irradiation unit 4 to the sample support unit 1 and the range of entrance of X-rays from the sample support unit 1 to the X-ray detector 5 overlap with each other. Moreover, the X-ray irradiation unit 4 and the X-ray detector 5 are located at symmetrical positions with respect to a hypothetical central axis, which is perpendicular to the plate-like sample support unit 1 and passes through the center of the through hole 11, and are located as proximally as practicable to the sample support unit 1. In FIG. 2, the hypothetical central axis is drawn with an alternate long and short dash line. The sample S is irradiated with X-rays from the X-ray irradiation unit 4, fluorescent X-rays are generated at the sample S, and the fluorescent X-rays are detected by the X-ray detector 5. In FIG. 2, X-rays from the X-ray irradiation unit 4 to be used for irradiation of the sample S and fluorescent X-rays, which are generated at the sample S and are detected by the X-ray detector 5, are drawn with dashed lines. The collimator 2 is located immediately below the sample support unit 1 and in an X-ray path from the X-ray irradiation unit 4 to the sample support unit 1. FIG. 3 is a schematic top plan view of the collimator 2. Actually, the sample support unit 1 is located further above the collimator 2. The collimator 2 is made of tantalum and is formed to have a plate-like shape, and a plurality of apertures 21, 22 and 23 having different diameters are formed at the collimator 2. The diameters of the apertures 21, 22 and 23 are, for example, 1.2 mm, 3 mm and 7 mm respectively. It is to be noted that the number of apertures is not limited to three but may be two or may be four or more. Moreover, there is no need to make all apertures have different diameters but it is only preferable to make at least one aperture have a diameter different from others. The apertures 21, 22 and 23 are aligned in a direction crossing a direction, in which the X-ray irradiation unit 4 and the X-ray detector 5 are aligned, in a horizontal plane. FIG. 4 is a schematic sectional view for illustrating the IV-IV cross section of FIG. 1. The plane illustrated in FIG. 4 corresponds to a plane perpendicular to the plane illustrated in FIG. 2. In FIG. 4, a hypothetical central axis passing through the center of the through hole 11 is drawn with an alternate long and short dash line. The X-ray detection apparatus is equipped with a linear drive motor 7 functioning as a movement mechanism of the collimator 2. The linear drive motor 7 is located at a position lower than the sample support unit 1. The linear drive motor 7 is equipped with a drive shaft 71, and the drive shaft 71 is connected with a parallel shaft 72 via a connecting plate 73. The linear drive motor 7 drives the drive shaft 71 linearly, and the parallel shaft 72 reciprocates in the longitudinal direction in conjunction with the drive shaft 71. The parallel shaft 72 is connected with the collimator 2. As the parallel shaft 72 reciprocates, the collimator 2 moves along an undersurface of the sample support unit 1. The collimator 2 can move in a direction, in which the apertures 21, 22 and 23 are aligned. The direction of movement corresponds to a direction perpendicular to the plane illustrated in FIG. 2, corresponds to the longitudinal direction of FIG. 3, and corresponds to the lateral direction of FIG. 4. As the collimator 2 moves, the apertures 21, 22 and 23 are shifted and any one of the apertures 21, 22 and 23 can be positioned in the X-ray path. When any one of the apertures 21, 22 and 23 is positioned in the X-ray path, X-rays pass through the one of the apertures 21, 22 and 23 and the sample S is irradiated with the X-rays from the X-ray irradiation unit 4. X-rays, which do not pass through the aperture, are blocked by the collimator 2. Illustrated in FIG. 3 is a state where the aperture 22 is positioned in the X-ray path. As the collimator 2 moves, an aperture through which X-rays pass is changed and the diameter of an aperture through which X-rays pass changes. As the diameter of an aperture changes, the size of X-rays to be used for irradiation of the sample S changes and the size of an analysis object part of the sample S changes. It is possible to select the size of an analysis object part of the sample S according to the objective, by selecting any one of the apertures 21, 22 and 23. The apertures 21, 22 and 23, which are located in the X-ray path from the X-ray irradiation unit 4, are positioned away from the hypothetical central axis passing through the center of the through hole 11. Moreover, the collimator 2 is connected with a window unit 62 which allows light to pass therethrough. The window unit 62 is constituted of a transparent member such as an acrylic plate. The window unit 62 is provided at a position along a direction in which the apertures 21, 22 and 23 are aligned. An image sensor 61 such as a CCD (Charge Coupled Device) image censor or a CMOS (Complementary Metal Oxide Semiconductor) image sensor is equipped immediately below the through hole 11. The X-ray detection apparatus is also equipped with light emitting 1 devices 63 and 63 such as Light Emitting Diodes. As the parallel shaft 72 reciprocates, the window unit 62 can be positioned immediately below the through hole 11. The light emitting devices 63 and 63 emit light when the window unit 62 is positioned immediately below the through hole 11, and the emitted light passes through the window unit 62, is reflected at the sample S, passes through the window unit 62 and enters the image sensor 61. In such a manner, the sample S is photographed. The X-ray detection apparatus is equipped with a control unit, which is not illustrated in the figures, for controlling the operations of the X-ray irradiation unit 4 and the linear drive motor 7. The control unit controls the operations of the linear drive motor 7 so as to control the position of the collimator 2. The control unit controls the position of the collimator 2 to locate the collimator 2 at one of a plurality of preset positions as needed. The plurality of positions for control include: positions to locate the apertures 21, 22 and 23 respectively in the X-ray path from the X-ray irradiation unit 4; and a position to locate the window unit 62 immediately below the through hole 11. The control unit allows X-ray irradiation by the X-ray irradiation unit 4 when any one of the apertures 21, 22 and 23 is positioned in the X-ray path from the X-ray irradiation unit 4. The control unit rejects X-ray irradiation by the X-ray irradiation unit 4 when the window unit 62 is positioned immediately below the through hole 11. It is to be noted that an X-ray detection apparatus may be constructed without including an image sensor 61, a window unit 62, and light emitting devices 63 and 63. The X-ray detection apparatus is also equipped with a signal processing unit, which is not illustrated in the figures, for executing signal processing for X-ray fluorescence measurement. The X-ray detector 5 outputs a signal proportional to the energy of detected fluorescent X-rays to the signal processing unit, and the signal processing unit executes processing to count signals of each value and obtain the relation between energy of fluorescent X-rays detected by the X-ray detector 5 and the number of counts, that is, a fluorescent X-ray spectrum. It is to be noted that the X-ray detector 5 may be constructed to detect fluorescent X-rays separately for each wavelength. Moreover, the signal processing unit may be constructed to further execute X-ray fluorescence analysis processing of making a qualitative analysis or a quantitative analysis of elements contained in a sample on the basis of the fluorescent X-ray spectrum. Furthermore, the collimator 2 is joined with the shield 3 for blocking X-rays. FIG. 5 is an enlarged view for illustrating the enlargement of a part including the collimator 2 and the shield 3 of FIG. 2. The shield 3 joined with the collimator 2 moves together with the collimator 2. The shield 3 is located at an intermediate position between the X-ray irradiation unit 4 and the X-ray detector 5 in a state where any one of the apertures 21, 22 and 23 is positioned in the X-ray path from the X-ray irradiation unit 4. In particular, in such a state, the shield 3 is located at an intermediate position on a line linking an arbitrary part of an exit 41 of X-rays at the X-ray irradiation unit 4 with an arbitrary part of an entrance 51 of fluorescent X-rays at the X-ray detector 5. As the shield 3 is located at the position, X-rays passing through a path liking the exit 41 with the entrance 51 is blocked, and X-rays are prevented from entering the X-ray detector 5 directly from the X-ray irradiation unit 4. The shield 3 is constructed by joining of a first shielding member 31 made of copper and a second shielding member 32 made of aluminum. The first shielding member 31 and the second shielding member 32 are separate bodies and are joined with each other using a screw or the like. A part, which is closer to the exit 41 of the X-ray irradiation unit 4, of parts of the shield 3 corresponds to the first shielding member 31, while a part, which is closer to the entrance 51 of the X-ray detector 5, of parts of the shield 3 corresponds to the second shielding member 32. The first shielding member 31 exists at an intermediate position on a line linking an arbitrary part of the exit 41 with an arbitrary part of the entrance 51, while the second shielding member 32 exists at an intermediate position on a line linking an arbitrary part of the first shielding member 31 and the collimator 2 with an arbitrary part of the entrance 51. The first shielding member 31 made of copper tends to absorb more intense X-rays than the second shielding member 32 made of aluminum. X-rays, which have exited from the exit 41 of the X-ray irradiation unit 4, are absorbed by the first shielding member 31. Although the absorption of X-rays causes generation of fluorescent X-rays of copper from the first shielding member 31, the fluorescent X-rays have lower intensity than the X-rays from the X-ray irradiation unit 4. The fluorescent X-rays generated from the first shielding member 31 are absorbed by the second shielding member 32. Although the second shielding member 32 also generates fluorescent X-rays, the fluorescent X-rays have further lower intensity and attenuate before entering the entrance 51. The effect of the fluorescent X-rays from the second shielding member 32 on X-ray fluorescence analysis is significantly smaller than that of the X-rays from the X-ray irradiation unit 4. Moreover, the shield 3 is provided at an intermediate position on a line linking an arbitrary part of the collimator 2 with an arbitrary part of the entrance 51. In a state where any one of the apertures 21, 22 and 23 is positioned in the X-ray path from the X-ray irradiation unit 4, the shield 3 exists between the entire collimator 2 and the entrance 51. The collimator 2 is irradiated with X-rays, which do not pass through the aperture, and fluorescent X-rays and scattered X-rays are generated from the collimator 2. To prevent the fluorescent X-rays and scattered X-rays from entering the X-ray detector 5, the first shielding member 31 blocks a part of X-rays generated from the collimator 2 and the second shielding member 32 blocks X-rays, which have not been blocked by the first shielding member 31, of the X-rays generated from the collimator 2. In such a manner, X-rays from the collimator 2 are blocked by the shield 3, and are prevented from entering the X-ray detector 5. Consequently, X-rays other than fluorescent X-rays generated from the sample S is prevented from entering the X-ray detector 5 effectively, and reduction in accuracy of X-ray fluorescence analysis is prevented. The X-ray detection apparatus can be used for highly accurate X-ray fluorescence analysis. The shield 3 is connected with an end, which is closer to the X-ray detector 5, of the plate-like collimator 2. Since the entrance 51 of the X-ray detector 5 is positioned closer to an undersurface than an upper surface of the collimator 2, the shield 3 is projected below the collimator 2 in order to block X-rays from the collimator 2 to the entrance 51. Moreover, the second shielding member 32 has a part projected above the collimator 2. Since the second shielding member 32 is projected to the upper and lower sides of the collimator 2, the shield 3 reliably blocks not only X-rays exiting from the collimator 2 to the lower side but also X-rays exiting from the collimator 2 to the upper side. This effectively prevents X-rays other than fluorescent X-rays generated from the sample S from entering the X-ray detector 5. The shield 3 is not provided below the window unit 62. The material of the collimator 2 is tantalum, which generates fluorescent X-rays having higher energy than copper that is the material of the first shielding member 31. Copper which is the material of the first shielding member 31 generates fluorescent X-rays having higher energy than aluminum which is the material of the second shielding member 32. In comparison with the energy of fluorescent X-rays generated from the collimator 2 which is mainly subjected to X-rays from the X-ray irradiation unit 4, the energy of fluorescent X-rays generated from the first shielding member 31 which is subjected to X-rays next is lower, and the energy of fluorescent X-rays generated from the second shielding member 32 is further lower. In a path of X-rays other than fluorescent X-rays from the sample S to enter the X-ray detector 5, energy of X-rays attenuates sequentially and X-rays are blocked effectively. It is to be noted that the materials of the collimator 2, the first shielding member 31 and the second shielding member 32 may be changed to other combinations, as long as fluorescent X-rays generated from the collimator 2 have the highest energy and fluorescent X-rays generated from the second shielding member 32 have the lowest energy. Moreover, an upper end part 33 of the second shielding member 32 has a shape not to block a path of fluorescent X-rays from the sample S to the entrance 51 of the X-ray detector 5. Specifically, the upper end part 33 has a shape to avoid a path linking an upper end of the through hole 11 with the entrance 51. Therefore fluorescent X-rays from the sample S enter the entrance 51 efficiently and are detected by the X-ray detector 5. Moreover, there is no need to provide a structure for fixing the shield 3 separately from the collimator 2, since the shield 3 is joined with the collimator 2 to move together with the collimator 2. This makes it possible to locate the shield 3 for blocking X-rays effectively even in a state where the X-ray irradiation unit 4, the X-ray detector 5, the sample support unit 1 and the collimator 2 are positioned as proximally as practicable to each other. Since it is possible to locate the shield 3 even in a downsized X-ray detection apparatus, it becomes possible to downsize an X-ray detection apparatus which can be used for highly accurate X-ray fluorescence analysis. FIG. 6 is a perspective view for illustrating the appearance of an X-ray detection apparatus. A main part of the X-ray detection apparatus is held in a case together with other parts, which are not illustrated in the figures, such as a power supply unit, and a part of the case is formed of a resin member 82. The case is covered with a cover part 81 made of metal. FIG. 7 is a perspective view for illustrating an X-ray detection apparatus in a state where the cover part 81 is open. The cover part 81 has a shape in which an end part of a plate is curved, and the curved end part functions as a movable end. An end part opposite to the curved end part functions as a connected end to be connected with the case by a hinge, and the cover part 81 can be opened and closed to the case by the operations of the hinge. In a state where the cover part 81 is closed, the movable end of the cover part 81 is fixed at the case with a stopper. The X-ray detection apparatus is provided with a plane part 83, which is constituted of an upper surface of the resin member 82. An upper side of the plane part 83 is covered with the cover part 81 in a state where the cover part 81 is closed, while an upper side of the plane part 83 is open in a state where the cover part 81 is opened. An opening is formed at the plane part 83, and the sample support unit 1 is located at the position of the opening. An upper surface of the detachable unit 12 is substantially in plane with a surface of the plane part 83. A main part of the X-ray detection apparatus other than the sample support unit 1 is located at a position below the plane part 83. FIG. 8 is a schematic sectional view for illustrating the VIII-VIII cross section of FIG. 6. A metal plate 85 is equipped at a position lower than the plane part 83. The metal plate 85 is located along the rear side of the resin member 82. The metal plate 85 has an opening at a part where the sample support unit 1 is located. A front end part of the metal plate 85 has a shape curving downward, and a back end part of the metal plate 85 has a shape curving upward. A shield box 84 made of metal is provided inside the case, and a main part of the X-ray detection apparatus other than the sample support unit 1, which is not illustrated in FIG. 8, is located inside the shield box 84. The upper side and the lateral sides of a main part of the X-ray detection apparatus other than the sample support unit 1 are surrounded by the shield box 84 made of metal. This makes X-rays to the upper side and to the lateral sides blocked by the shield box 84. X-rays which have leaked to the upper side of the sample support unit 1 are blocked by the cover part 81 made of metal. Moreover, X-rays radiated from the upper side of the sample support unit 1 to the front direction of the X-ray detection apparatus are blocked by end parts of the cover part 81 and the metal plate 85. For closing the cover part 81, the movable end of the cover part 81 overlaps with the metal plate 85 in the front direction. This makes X-rays radiated from the upper side of the sample support unit 1 to the front direction of the X-ray detection apparatus collide with any one of end parts of the cover part 81 or the metal plate 85 and blocked reliably. With such a structure, the X-ray detection apparatus can block X-rays reliably and ensure safety even though the case includes the resin member 82. Since a part of the case is made of resin, weight saving of an X-ray detection apparatus can be realized. Weight saving of an X-ray detection apparatus can improve the portability and the convenience of the X-ray detection apparatus. As this invention may be embodied in several forms without departing from the spirit of essential characteristics thereof, the present embodiment is therefore illustrative and not restrictive, since the scope of the invention is defined by the appended claims rather than by the description preceding them, and all changes that fall within metes and bounds of the claims, or equivalence of such metes and bounds thereof are therefore intended to be embraced by the claims. |
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043953816 | claims | 1. A confinement enclosure for a nuclear reactor, comprising: a housing in which a nuclear reactor is located, said housing consisting of a single wall formed of reinforced concrete; containment means in said single wall for preventing flow of gas through said wall in the event of an accident involving the nuclear reactor, said containment means including a network of channels defined in said wall; filter means located within said housing and fluidly connected to said channels to filter any fluid entering said channels from within said housing; and pressurizing means for connecting said filter means to said containment means, said pressurizing means being fluidly connected to said channels and at all times maintaining a pressure on filtered fluid inside said channels at a level which is higher than the pressure inside said housing and higher than any fluid by-passing the filter means and passing through a crack in said wall; whereby any fluid by-passing the filter means is prevented from entering any channel and only fluid passing through said filter means into said channels is permitted to reach the outside of said housing. a housing in which a nuclear reactor is located, said housing consisting of a single wall formed of reinforced concrete; containment means in said single wall for preventing flow of gas through said wall in the event of an accident involving the nuclear reactor, said containment means including a network of channels defined in said wall; pressurizing means fluidly connected to said channels and at all times maintaining a pressure on fluid inside said channels at a level which is higher than the pressure inside said housing and higher than any fluid passing through a crack in said wall, said channels being located in a plurality of zones with each zone corresponding to a section of wall height, said channels being grouped according to said zones, and the channels of each zone being filled with water in communication with said pressurizing means constituted by a free level water reservoir located inside the enclosure at a higher level than the corresponding height section; whereby any fluid passing through a crack in said wall is prevented from entering any channel and only fluid passing into said channels is permitted to reach the outside of said housing. 2. A confinement enclosure for a nuclear reactor, comprising: 3. A confinement enclosure according to claim 2, wherein said housing has a bottom provided with a drainage network filled with water from a free level reservoir inside the enclosure, the pressure in said drainage network being established by said reservoir. |
050698273 | abstract | A process for reducing and dissolving PuO.sub.2 comprises contacting the plutonium dioxide with a hydrazine-free, acid aqueous solution containing a reducing agent, whose redox potential is below +0.5 V/ENH in order to reduce and dissolve the plutonium. The reducing agent can be Cr.sup.2+ or U.sup.4+. |
042882923 | claims | 1. Apparatus for providing connection between a first member and a second member movable within a horizontal plane with respect to said first member, comprising: a. a first plurality of affixed vertical surfaces laterally spaced, each having a laterally extending upper restraint; b. a second plurality of affixed vertical surfaces laterally spaced, each having a laterally extending lower restraint; c. structure for affixing said first surfaces to said first member and said second surfaces to said second member such that said first surfaces are laterally aligned with said second surfaces and are vertically spaced therefrom a preselected distance; d. a plurality of flexible cables of substantially similar diameter, each said cable being stationary at one portion with respect to said first member and stationary at another portion with respect to said second member and positioned partially between two consecutive first surfaces and partially between two consecutive second surfaces and between their respective restraints so as to form a generally C-shaped loop in a substantially vertical orientation which slidingly moves with respect to said surfaces upon relative motion of said first and second members; and e. a plurality of flexible devices affixed to said first surfaces for supporting the upper portion of each said looped cable, said flexible device including a flexible body affixed to one of said first vertical surfaces having an upper section and a crimped lower segment, said lower segment spaced from one of said vertical surfaces a distance less than said diameter and said upper section spaced from said last-named surface a distance greater than said diameter, whereby upon relative motion of said members in one direction said loops of said cables contact said crimped segment from below, flex said segment laterally, and pass above said segment, and upon relative motion in the opposite direction said loops flex said segment laterally and pass free from said segment. a. a first plurality of affixed vertical surfaces laterally spaced, each having a laterally extending upper restraint; b. a second plurality of affixed vertical surfaces laterally spaced, each having a laterally extending lower restraint; c. structure for affixing said first surfaces to one of said plug and said one stationary member and said second surfaces to the other of said plug and said stationary member such that said first surfaces are laterally aligned with said second surfaces and are vertically spaced therefrom a preselected distance; d. a plurality of flexible cables of substantially similar diameter, each said cable being stationary at one portion with respect to said first surfaces and stationary at another portion with respect to said second surfaces and positioned partially between two consecutive first surfaces and partially between two consecutive second surfaces and between their respective restraints so as to form a generally C-shaped loop in a substantially vertical orientation which slidingly moves with respect to said surfaces upon relative motion of said first and second members; and e. a plurality of flexible devices affixed to said first surfaces for supporting the upper portion of each said looped cable, said flexible device including a flexible body affixed to one of said first vertical surfaces having an upper section and a crimped lower segment, said lower segment spaced from one of said vertical surfaces a distance less than said diameter and said upper section spaced from said last-named surface a distance greater than said diameter, whereby upon relative motion of said members in one direction said loops of said cables contact said crimped segment from below, flex said segment laterally, and pass above said segment, and upon relative motion in the opposite direction said loops flex said segment laterally and pass free from said segment. 2. A nuclear reactor including substantially stationary members, a horizontally rotatable vessel head plug mounted above a nuclear core, and apparatus for providing connection between one of said members and said plug comprising: 3. The apparatus of claim 2 wherein at least one of said cables is adjacent another one of said cables such that the loop of said one cable is of smaller radius than the loop of said adjacent cable and is contained within the substantially vertical area enclosed by the loop of said another cable. 4. The apparatus of claim 3 wherein all of said cables have a substantially similar outside diameter and wherein the lateral spacing between consecutive vertical surfaces is approximately 11/2 said diameter. 5. The apparatus of claim 3 wherein all of said cables have a substantially similar outside diameter and said preselected distance is less than said diameter. |
053613775 | abstract | A method for producing electrical power from steam generated by a nuclear reactor comprising the steps of: providing a nuclear reactor engaged to a steam generator for generating steam when heated aqueous product is passed therethrough; and passing heated aqueous product through the steam generator to produce steam. The method additionally comprises passing the produced steam through a fossil fired or steam to steam superheater to superheat the produced steam; and passing the superheated produced steam through a first turbine to expand the superheated produced steam and produce steam. The produced steam from the first turbine is subsequently reheated to obtain a reheated steam. The obtained reheated steam is then passed through a second turbine coupled to a generator in order to expand the obtained reheated steam and generate electrical power with the generator. An apparatus to accomplish the method for producing electrical power from steam generated by a nuclear reactor. |
052572986 | claims | 1. Nuclear fuel pellets including a fission substance of UO.sub.2, the nuclear fuel pellets comprising UO.sub.2 grains and an aluminosilicate deposition phase, the deposition phase being a glassy state, a crystalline state or a combination thereof, said grains having an average grain diameter of about 20 .mu.m through about 60 .mu.m, said aluminosilicate deposition phase having a composition including SiO.sub.2 of about 60 wt % through about 80 wt % and Al.sub.2 O.sub.3 of the residual on average, the amount of alumina plus silica being about 10 ppm through about 500 ppm with respect to the total amount of said nuclear fuel pellets, said pellets having a porosity of 5 vol % at a maximum. 2. The nuclear fuel pellets of claim 1, wherein the amount of alumina plus silica is about 10 ppm through 30 ppm with respect to the total amount of said nuclear fuel pellets. 3. The nuclear fuel pellets of claim 1, wherein the amount of alumina plus silica is about 10 ppm through 70 ppm with respect to the total amount of said nuclear fuel pellets. 4. The nuclear fuel pellets of claim 1, wherein the amount of alumina plus silica is about 10 ppm through 130 ppm with respect to the total amount of said nuclear fuel pellets. 5. The nuclear fuel pellets of claim 1, wherein the amount of alumina plus silica is about 10 ppm through about 250 ppm with respect to the total amount of said nuclear fuel pellets. 6. Nuclear fuel pellets including a fission substance of UO.sub.2 having Gd.sub.2 O.sub.3 added thereto, the nuclear fuel pellets comprising (U, Gd) O.sub.2 grains and an aluminosilicate deposition phase, the deposition phase being a glassy state, a crystalline state or a combination thereof, said grains having an average grain diameter of about 20 .mu.m through about 60 .mu.m, said aluminosilicate deposition phase having a composition including SiO.sub.2 of about 40 wt % through about 80 wt % and Al.sub.2 O.sub.3 of the residual on average, the amount of alumina plus silica being about 10 ppm through about 500 ppm with respect to the total amount of said nuclear fuel pellets, said pellets having a porosity of 5 vol % at a maximum. 7. The nuclear fuel pellets of claim 6, wherein the amount of alumina plus silica is about 10 ppm through 30 ppm with respect to the total amount of said nuclear fuel pellets. 8. The nuclear fuel pellets of claim 6, wherein the amount of alumina plus silica is about 10 ppm through 70 ppm with respect to the total amount of said nuclear fuel pellets. 9. The nuclear fuel pellets of claim 6, wherein the amount of alumina plus silica is about 8 ppm through 130 ppm with respect to the total amount of said nuclear fuel pellets. 10. The nuclear fuel pellets of claim 6, wherein the amount of alumina plus silica is about 10 ppm through about 250 ppm with respect to the total amount of said nuclear fuel pellets. |
048760560 | description | The apparatus 10 includes an irradiation chamber 16, cylindrical in shape, made of steel coated with an anticorrosion plastic coating, and with a small vent hole 18 in its top. The chamber 16 encloses a copper tube 20 wound into a number of helically wound layers; the portion of the tube 20 forming the innermost layer is of narrower diameter than the remainder of the tube 20. An end portion 22 of the tube 20 extends from the innermost layer of the winding out of the base of the chamber 16 and is connected via a constant volume, high pressure, injection pump 24 to the bore of the pipeline 12. The other end portion 26 of the tube 20 projects from the chamber 16, tapering to a narrower diameter, and being open at its end 28. Since the apparatus 10 is immersed in sea water, sea water fills the chamber 16 and the entire length of the tube 20; the volume of sea water in the tube 20 is about 50 liters. At the centre of the irradiation chamber 16 is a 100 microgram californium-252 neutron source 30 providing 2.3.times.10.sup.8 neutrons/sec. These are moderated by the sea water so the sea water in the tube 20 is subjected to irradiation by thermal neutrons. At a distance of 40 mm from the source 30 the flux is about 10.sup.6 thermal neutrons per cm.sup.2 per second. The pump 24 is connected by a power supply cable 32 to a microprocessor unit 34. The unit 34 is connected by a cable 36 to a sodium iodide scintillator 38 mounted on the wall of the pipeline 12 at a distance in the flow direction from the point at which the injection pump 24 is connected to the pipeline 12 such that the oil/water in the pipeline 12 takes about 10 seconds to travel that distance. The scintillator 38 itself is of such a length that the oil/water in the pipeline 12 flows past it in about a tenth of a second. The unit 34 is also connected by a power-supply and data transmission cable 40 to a power source (not shown) which may be remote from the apparatus 10. In operation of the apparatus 10 the sea water filling the tube 20 is constantly irradiated by thermal neutrons from the source 30, so that sodium-24 nuclides are generated from the sodium naturally present in the sea water. Once every fifteen minutes the pump 24 is energised to inject 50 milliliters of this irradiated sea water into the pipeline 12. After about ten seconds this injected sample of sea water will pass the scintillator 38 which detects the gamma radiation emitted as a result of the decay of the sodium-24 nuclides. From the time interval between energising the pump 24 and the detection of the gamma rays by the scintillator 38, the flow velocity of the contents of the pipeline 12 are calculated by the microprocessor 34, this data being transmitted along the cable 40. Since the end 28 of the tube 20 is open, the tube 20 remains full of sea water at all times, and the sea water flows along the tube 20 slowly (at a rate corresponding to 50 ml per 15 minutes) taking just over ten days to flow the length of the tube 20. The long helical winding and the narrow end portion 26 minimise diffusion of irradiated sea water back out of the open end 28. Assuming for simplicity that the neutron flux is uniform throughout the length of the tube 20, and considering a small volume of the sea water in the tube 20 which at a time t includes N.sub.S stable atoms and N.sub.R radioactive atoms (whose sum is N.sub.T). Then: EQU dN.sub.R =FN.sub.S dt-cN.sub.R dt where F is a constnt depending on the neutron flux and the cross-section for production of the radioactive nuclei, and where c is the decay constant for the radioactive nuclei. Consequently, after irradiation of the sea water for a time t.sub.r : EQU N.sub.R =(FN.sub.T /c)(1-e.sup.-ct.sbsp.r) is the number of radioactive atoms present. (This assumes that only a small proportion of the atoms present become radioactive). This small volume of seawater is then injected into the pipeline 12, and passes along the pipeline 12 past the scintillator 38. If the transit time taken for it to reach the scintillator 38 is t.sub.t, and the time during which gamma rays from the irradiated sea water are counted is t.sub.c, then the number of gamma rays detected will be: EQU n=(KFN.sub.T /c)(1-e.sup.-ct.sbsp.c)e.sup.-ct.sbsp.t (1-e.sup.-ct.sbsp.r) where K is a constant. Since in practice the counting time will be much shorter than the transit time and the irradiation time, and will be much less than the half-life of the radioactive nuclei (i.e. since ct.sub.c is much less than one) this can be written as: EQU n=KFN.sub.T t.sub.c e.sup.-ct.sbsp.t (1-e.sup.-ct.sbsp.r) It is evident that the number of gamma rays detected depends in general upon the decay constant c. For any given values of irradiation time and transit time there is an optimum value of c which gives the largest number of detected gamma rays. This corresponds to an optimum half-life T lying between the values of the transit time and the irradiation time, and given by: EQU T=t.sub.r log 2/log ((t.sub.r +t.sub.t)/t.sub.t) For example in the apparatus 10 the sample of sea water is irradiated for ten days before injection, and the transit time is about 10 seconds. For these values, the optimum half-life of the radioactive nuclei is: EQU T=52 700 sec=14.6 hours This is close to the half-life of sodium-24, which is 14.8 hours. However, as long as the irradiation time is several half-lives, then the equilibrium concentration of the radioactive nuclide will be achieved; and as long as the transit time is a small fraction of the half-life, then the activity will not significantly decrease during transit. Thus the values of the irradiation time, half-life, and transit time are not critical. Under such circumstances the decay constant c does not affect the number of gamma rays detected, which is given by: EQU n=KFN.sub.T t.sub.c and hence the factors which determine the number of gamma rays detected are the parameters F, relating to the neutron flux to which the test fluid is subjected, and to the cross-section for the activation reaction, and N.sub.T relating to the concentration in the test fluid of the nuclides which can undergo the activation reaction, and to the volume of the injected sample. In the present case, the cross-section of sodium is 0.54 barns; and as mentioned earlier the concentration of sodium ions typically 10 g/liter. If any other gamma-emitting nuclides are also generated (for example a chlorine isotope) this will increase the number of gamma rays detected. It will be appreciated that the dimensions of the tube 20 may differ from those described, as too may the volume of the injected sample. The injected volume is chosen to ensure that passage of the injected sample can be reliably sensed by the scintillator 38, i.e. that sufficient gamma counts are received; and that the gamma counts received form a sharp peak, i.e. that the injected test fluid occupies only a short length within the pipeline 12. In the apparatus 10 it will be appreciated that the detected gamma count can be increased by increasing the injected volume of test fluid; although this reduces the irradiation time, and hence the activity of the test fluid, this is more than made up for by the increased volume. However the injected volume must be small in comparison with the volume per unit length of the pipeline 12, and the time taken for the injection must be short in comparison with the transit time. |
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abstract | An apparatus includes a radiation source that emits a radiation beam that causes substantially all of a quantity of material to evaporate; and structure having first and second surface portions, a first operational mode wherein a greater quantity of a byproduct of the evaporation impinges on the first surface portion, and a second operational mode wherein a greater quantity of the byproduct impinges on the second surface portion. A different aspect involves emitting a radiation beam toward a quantity of material, the radiation beam causing substantially all of the quantity of material to evaporate; operating a structure having first and second surface portions in a first operational mode wherein a greater quantity of a byproduct of the evaporation impinges on the first surface portion; and thereafter operating the structure in a second operational mode wherein a greater quantity of the byproduct impinges on the second surface portion. |
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description | The basic design of the spherical container and of its shell are described with general reference to the drawings and particular reference to FIG. 1. The spherical container is denoted as a whole by the reference symbol 10. It has a spherical shell 12 with a preferably uniform thickness D. The spherical shell consists of a plurality of layers in the radial direction. The innermost layer 14 is a layer made from a material which emits as few particles as possible in a temperature range as wide as possible. Furthermore, the inner layer must allow the force field generated by a second spherical shell layer to pass into the cavity of the interior of the sphere. In the exemplary embodiment shown, a second layer 16, the so-called magnetic layer, generates a magnetic field in the interior of the sphere which is designed such that anti-matter, if it is magnetically or electrically charged, is kept in the interior of the sphere without coming into contact with the wall. The second layer is expediently built up from individual, juxtaposed spherical shell segments 18 which are electrically controlled individually or in groups of suitable number by a central control device which ensures, for its part, that the field in the interior is uniform. The ferromagnetic anti-matter, for example anti-iron, is then uniformly attracted by the magnets in the envelope of the sphere. This produces at the center of the sphere and around a spherically symmetrical center domain a site for the anti-matter at which it can be kept in a state of equilibrium by suitable open-loop and/or closed-loop control operations, so that it does not come into contact with matter. Viewed in the radial direction, the individual magnets filling a spherical shell segment 18 are always identically polarized, for example the south pole can always be directed into the middle. Also provided is an evacuation opening 22 which radially penetrates the spherical shell and is connected to a corresponding vacuum generator. An outer layer 17 of the spherical shell consists of a material which is matched to the location and application of the storage container. Depending on the requirements made, this could, for example, be ceramic, in order to withstand high temperatures and offer only low heat conductivity toward the inside, or else it could also be a high-grade steel in order to ensure high mechanical loadability. Further exemplary embodiments of the materials for the outer layer can be put together in a corresponding way. Layers 14, 16, 17 are secured against mutual rotation by suitable projections and indentations in which these projections engage mechanically. However, it is also possible to use other anti-rotation means, depending on the material of the layers and the current application of the container. Electrically charged anti-matter is likewise kept in the center of the sphere by the magnetic field. With reference to FIG. 2, a further exemplary embodiment of the present invention will now be explained which can be used in conjunction with the spherical shell illustrated in FIG. 1 and a magnetic field in the interior of the sphere, but is likewise the subject matter for a particularly preferred exemplary embodiment in which electrically charged anti-matter is kept in the interior of the sphere by an electrostatic field. This latter embodiment is treated further below. In accordance with FIG. 2, the spherical container 10 is provided with a filling and emptying device which docks on the right-hand side in FIG. 2 or the left-hand side on the spherical container 10. The filling and emptying device is provided as a whole with the reference symbol 30. It comprises a continuous frame 32 which mechanically connects the part 34 illustrated on the right in the figure with the part 36 illustrated on the left in the figure. The prime task of the frame is to lend the arrangement 30 an appropriate mechanical stability, and to ensure that the arrangement 30 can be guided stably in a fashion suiting the openings located in the spherical shell, specifically a first, larger opening 38 and a second, smaller opening 40. The left-hand region 36 of the filling and emptying arrangement is formed by a cylindrical attachment 42. On the side directed toward the spherical shell, it has an opening which surrounds the first, larger opening 38 in the spherical shell, such that when properly used the opening cross section of the latter is preferably not reduced by the attachment 42. Arranged in the interior of the attachment 42 are two closing pieces, a first closing piece 44, which fits the larger opening 38 in the spherical shell, and a second closing piece 46, which fits the second, smaller opening 40 in the spherical shell. Like the remainder of the spherical shell, the closing pieces 44 and 46 comprise the above-described spherical shell segments. Starting from their convex outer surfaces, both closing pieces are coupled by means of a rigid connection, a rigid frame rod 47 which can absorb tensile and thrust forces in the direction of the illustrated longitudinal axis of the cylindrical attachment, and which ensures that the closing pieces can move as pairs with the same distance between them when a force is-exerted on one of the two closing pieces. This movement is then guided by the rigid coupling in such a way that it extends exactly in the axial direction through the center of the sphere and along the axis of the cylindrical attachment 42. By means of a drive represented only diagrammatically in the illustration, the two closing pieces can be moved along the central axis of the cylindrical attachment 42 out of the position illustrated in FIG. 2 and through the opening 38, and be guided further in this direction until the closing piece 46 fits in the opening 40 and, at the same time, the closing piece 44 fits in the opening 38. The lateral surface of the cylindrical attachment 42 has an inner layer structure which is similar to that of the spherical shell, in order likewise to be able to exert forces of magnetic attraction on the anti-matter in the interior thereof. The same reference symbols therefore apply here. Also located in the lateral surface of the cylindrical attachment is an introduction channel for anti-matter, which is re-machined in the manner of the design, mentioned in the introduction to the description, of the tubular storage container for anti-matter described by NASA. Reference is therefore made to the above-mentioned publication for further details on this introduction channel. The introduction channel opens into the cylindrical attachment 42 through an opening 52. The filling of the spherical container 10 with anti-matter which is ferromagnetic, that is to say, for example, anti-iron, is described below with reference to FIG. 2. The anti-matter passes through the introduction channel 50 into the interior of the cylindrical attachment 42. In order to control its position and movement in such a way that it does not come into contact with any wall bordering on the interior of the attachment 42 or of the shell 10, its movement is controlled by essentially three arrangements A, B and C which each generate a magnetic field which permits the above-described movement of the anti-matter. A and B are located on mutually opposite sides of the introduction channel and generate magnetic fields which are directed upstream. The direction of flow is illustrated in FIG. 2 by an arrow. A and B serve to brake the movement of the anti-matter. The movement of the matter in the direction of flow is rendered possible by a magnetic field C which is arranged opposite the introduction opening 52, and by corresponding magnets in the wall region, at that location, of the cylindrical attachment 42. Depending on the speed of the anti-matter, the polarity of C can be reduced, or even inverted. It is therefore possible to prevent the anti-matter from impinging in the region of the wall in which the magnetic field C is generated. The anti-matter is now located as a certain, prescribed quantity in a spherical mass X at the point of intersection between the axis of the introduction channel 50 and that of the cylindrical attachment 42. The introduction channel is now closed. An appropriate device is provided for this purpose. How the anti-matter reaches into the interior of the sphere will be described below. The direction of magnetization of the magnets contained in the closing pieces 44 and 46 can be controlled starting from their polarity and from the strength of the magnetic field. The closing pieces are now guided, as described above, into the interior of the spherical container by means of the above-mentioned mechanical drive, which can likewise be matched to the respective location and application of the overall device. During this movement, there builds up between the two closing pieces 44 and 46 a magnetic field which has a gradient which is set precisely so fine that during this movement the anti-matter is always located at the midpoint between the two closing pieces. The fine tuning of the magnetic fields D and E is performed by a central controller which is computer-based and whose most important input parameters are the speed/time characteristic of the closing pieces, the mass of anti-matter obtained, and the distance of the closing pieces from one another. The position drawn in FIG. 3 is reached in this way in the course of the movement, and at the end of the movement it is the closed position reached in FIG. 4. At the end of this movement, the magnetic field of the closing pieces D and E is controlled such that, the closer they come to their closed end position, the magnetic field strength approaches that field strength of the respectively neighboring spherical shell segments. This ensures that the anti-matter passes safely into the interior of the sphere. As described above, they are then located there in a position of equilibrium which can be maintained by exact open-loop control of the magnetic fields of the individual magnet segments. The maintenance of the anti-matter in the center of the spherical container 10 can advantageously be performed by closed-loop control by means of a closed control loop in addition to open-loop control. The feedback signal required for feeding the positional information back into the control loop can be generated using the most varied sensors. By virtue of the fact that the individual magnets which are contained in the spherical shell segments 18 can be set individually in terms of the strength of their magnetic field, it is possible for the magnetic field to be strengthened or weakened in one or other direction. Consequently, to the extent that it can be attracted by magnetic forces, the anti-matter moves in the interior of the spherical container 10. When dimensioning the size of the individual magnets in the spherical shell segments 18, it should be ensured that the temporal inertia of changes in magnetic field which the individual magnets experience for the purpose of the closed-loop control is as low as possible, in order to achieve efficient closed-loop control. For example, one or more light beams could be interrupted by the anti-matter when the latter is located in the center of the spherical container. The state of the interrupted light beams could then be converted into a signal which, within certain, prescribed tolerances, signals correct positioning of the anti-matter with reference to a plane in space. A repetition of the same principle for various other planes in space thus permits three-dimensional closed-loop control of the storage location of the anti-matter in the spherical container 10. With reference to FIG. 5, a description is given below of a further, preferred exemplary embodiment of the anti-matter storage device according to the invention which can be applied whenever the anti-matter is electrically charged. In this case, the same reference symbols denote the same parts as in FIGS. 2, 3 and 4. In a departure from the above-described exemplary embodiment, the shell of the spherical container 10 is now designed in principle as a spherical capacitor. Consequently, the materials of the individual layers of the layer-type composition of the spherical container must be matched in terms of the material. An embodiment of the spherical container which is suitable for negatively charged anti-matter is described below. It can be altered for anti-matter with opposite charge by appropriate modification. The negatively charged anti-matter can be produced, in principle, by bombardment with electrons which, for their part, utilize the valence positrons of the respective molecules. The shell, designed as a spherical capacitor, of the spherical container 10 comprises the three electrically required layers, specifically an inner capacitor spherical shell 60, which is negatively charged, a dielectric 62 and-an outer capacitor spherical shell 64 which is positively charged. These electrically active shells are preferably arranged concentrically. The shells are arranged secured against mutual rotation with the aid of anti-rotation means such as have already been explained in principle in the preceding exemplary embodiment. The anti-rotation means are not permitted to conduct any electric current. Electric charge is provided to outer shell 64 by a charging circuit 70 of any suitable or desired construction. Charging circuit 70 is controlled by a control device 74. A second charging circuit 72 may also be provided so that charge can be distributed separately and selectively transported to either of the upper one and a lower hemisphere of shell 64. The layer thickness, the distance and the materials must be selected as a function of the strength of electric field which is to be built up in the interior of the sphere. Once the anti-matter is centered in the interior of the spherical container, it is repelled uniformly from all sides, since the charges of anti-matter and of the inner spherical wall are equal. In the present case, both are negatively charged. It is thereby possible to dispense with complicated closed-loop control of the position of the anti-matter, since the anti-matter automatically takes up position in the center of the interior of the sphere, since in the absence of a gravitational field, producing a weight of the anti-matter, this site is the site of lowest potential energy. This site is displaced in the direction of the gravitational field when such a field is present. The closed-loop control of the electric field strength of the electrostatic field is therefore performed in a preferred way such that the weight of the anti-matter is compensated in an appropriate way by the strength of the electric field so that the matter does not come into contact with the wall. If the storage device is fixed at a set orientation relative to the direction of the gravitational field, the spherical shell and, similarly, a half-cylinder shell can generate a larger repulsive force than the respective other one. For this purpose, the half shells are then isolated electrically from one another and can be electrically charged using separate control circuits. The anti-matter can now in principle be conveyed into the sphere, or out of it, in the same way as was explained in the preceding exemplary embodiment. In the present case, the inner wall of the introduction channel 50 is likewise negatively charged, in order to prevent the negatively charged anti-matter from coming into contact with its walls. The inner wall of the cylindrical attachment 42, and the inner walls of the closing pieces 44 and 46, are also negatively charged for the same reason. The gradient of the electric field, which is required for the purpose of introducing the anti-matter in this exemplary embodiment, can be instituted by a weakening of the repulsive electric field which is generated in the closing piece 46. Likewise, a strengthening of the electric field in the closing piece 44 can be instituted. It is also possible to combine both these measures. The inner capacitor spherical shell consists of metal it the present exemplary embodiment. The dielectric intermediate shell expediently consists of a material which is best suited to the application and location as well as to the physical conditions occurring there such as temperature, pressure etc. The outer capacitor shell likewise consists of metal. In principle, the two closing pieces 44 and 46 have the same capacitor design as the remainder of the spherical shell. The open-loop control of the electric field in the interior of the sphere can be instituted from outside by appropriate supply leads. In a modification of-the last-described exemplary embodiment, it is also possible to dispense with the outer, positively charged layer, since it is, after all, only the radially inwardly directed, electrostatic field generated by the inner spherical shell which is relevant for stable storage of the anti-matter. The emptying operation can be implemented by correspondingly reversing the steps required for filling. Although the present invention was described above with the aid of a preferred exemplary embodiment, it is not limited thereto, but can be modified in multifarious ways. All these modifications are to be covered by the scope of protection of the claims as these are specified below. Spherical container Inner layer Second layer Outer layer Segments Evacuation opening Filling and emptying arrangement Frame Right-hand part of the frame Left-hand part of the frame First large opening Second small opening Cylindrical attachment First closing piece Frame rod Second closing piece Introduction channel Introduction opening Inner capacitor spherical shell Dielectric Outer capacitor spherical shell |
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abstract | To provide a containment cask for storage or transport of radioactive material, without employing a homogenization treatment. Pouring a molten lead between an inner shell 2 and an intermediate shell 3 to serve as a gamma ray shielding material, allowing the lead to cool, and subsequently, filling either one or both of a first void layer 9a formed at a boundary between the inner shell 2 and the poured lead 5a or a second void layer 9b formed at a boundary between the intermediate shell 3 and the poured lead 5a, using a low melting point metal 10 in a closely adhering state. To provide the cask 1 with a good heat-dissipating effect, by filling the void layers 9a, 9b that prevent the cask 1 from dissipating heat, with the low melting point metal 10 that has a superb thermal conductivity. |
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063317125 | summary | BACKGROUND OF THE INVENTION The present invention relates to a method for forming a section of a sample, such as a semiconductor integrated circuit and ceramics substrate, having a plurality of conductive layers by using a focused ion beam to observe the section. There is known, as a method to form and observe a section for observing a fine structure of a sample, a method of repeatedly irradiating a focused ion beam to a region having as a side a section observation position of a sample to form a recess (hole) in the sample as disclosed for example in JP-A-2-123749 so that a section of the sample appeared in a sidewall of the recess is observed by scanning and irradiating another charged particular beam. FIGS. 1A and 1B show, as an example of a sample including a plurality of conductive layers, an interconnect leading to a gate of a MOSFET. Explanation will be made on a method to observe a section obtained by cutting the interconnect. FIG. 1A is a plan view of a gate interconnect 51 portion. A source region 52 and a drain region 53 are formed on both sides of the gate interconnect 51 in a surface portion of a substrate 50. The source interconnect 55 and the drain interconnect 56 are respectively connected to the source region 52 and the drain region 53 through contacts 54. Explanation is made for a case that the gate interconnect 51 is observed, from a direction of an arrow C, in a section at a position (broken line A-B) between the source region 52 and the drain region 53. A work frame 60 (hole) is defined as in FIG. 1A, which is a region where a focused ion beam is irradiated to form a recess. Subsequently, a focused ion beam is scanned within the work frame 60 to open a hole. As a result, a recess 61 is formed as shown in FIG. 1B. The section of the gate interconnect 51 can be observed by observing in a direction of an arrow D shown in FIG. 1B. At this time, the gate interconnect 51 (on a D side) leading to the MOSFET gate is separated from the conductive layers, such as other interconnects and semiconductor substrate, by the section formation, and is thus rendered in an electrically floating state. Because the floating-state gate interconnect 51 leading to the gate is formed on a substrate 50 surface through a gate oxide film 57, they form a capacitor. In this state, if charged particles, such as a focused ion beam or electron beam, are irradiated for observation, the charges are charged in the capacitor thus formed. A so-called charge-up phenomenon occurs wherein the potential of the floating gate interconnect 51 increases with respect to a potential of the substrate 50 when irradiating a focused ion beam, and decreases in the case of an electron beam. That is, a problem occurs because of the charge-up phenomenon when the section is observed by irradiating with a focused ion beam or electron beam in a D direction. In particular, generally in the case of an ion beam the effected charge-up is prominent. That is, in the case of a focused ion beam, the secondary electrons are withdrawn in the section due to charge-up in the observation surface (interconnect section) to a positive potential. This makes it impossible to detect sufficient secondary electrons for obtaining an image. As a result, there is a problem in that the interconnect rendered in charge-up for the image is dark and observation is impossible. As a countermeasure to this, there is a method disclosed for example in JP-A-7-45681 that the conductive layer in electrically floating is formed with a section separately from one for observation, an exposed new section is irradiated by a focused ion beam while blowing a metal compound gas thereby forming a metal film, and the conductive layer in floating is electrically connected to a sample substrate to avoid a charge-up. This method is effective as a means to avoid a charge-up phenomenon. However, there have been such problems that a metal compound gas blowing mechanism is required besides the focused ion beam irradiation system, and there is necessity of operations of second section formation and the succeeding metal film formation. Meanwhile, in recent years line widths have become narrow and the gate oxide film formed between the substrate 50 and the gate interconnect 51 has become thin. In particular, the area of the floating conductive layer has been decreased by section formation due to refining in a forming region. Due to this, there has been a drastic decrease in capacitance of a capacitor formed in the process of section formation, and there has been increase in relative potential difference of the gate to the substrate due to floating phenomenon. As stated before, because the decrease in gate oxide film thickness results in a lower withstand voltage, the possibility of sample damage due to charge-up is increased. In particular, before forming a floating conductive layer in the process of section formation, the electric charges due to focused ion beam irradiating are discharged through the conductive layer. The section is observed by irradiating a scanning focused ion beam or scanning electron beam to the formed section and detecting secondary charged particles occurring from the section. However, after a recess 61 for section observation is formed and the conductive layer is rendered in a floating state, electric charges are accumulated in the floating conductive layer during formation of a portion lower than the conductive layer, causing a charge up phenomenon. There is a problem in that a high voltage is applied to the thin gate oxide film that is low in withstand voltage, resulting in damage to the sample as the case may be. SUMMARY OF THE INVENTION The present invention has an object to solve these problems. That is, in order to avoid the problem due to the floating conductive film, a second conductivity layer region formed in a electrically floating state by the section forming process is electrically connected to another non-floating first conductivity layer region (substrate). At this time, realization is made by continuously irradiating a focused ion beam to one point without scanning, or otherwise irradiating it with scanning over an extremely narrower region than the recess formed for section observation. That is, the present invention is a sample section formation observing method, in a sample including a second conductive layer overlying a first conductive layer through an insulating layer, characterized by comprising: a first process of irradiating a focused ion beam repeatedly scanning at a predetermined region of the sample and forming a recess to form a sidewall in which the second conductive layer and at lease the first conductive layer are exposed, a second process of irradiating the focused ion beam at a region other than the predetermined region from above the second conductive layer rendered in an electrically floating state by the first process to provide a hole reaching the first conductive layer thereby electrically connecting between the first conductive layer and the second conductive layer, and a third process of observing a desired exposed portion in the recess using a charged particle beam. Although scan irradiation is best effective at a site where the second conductive layer in a floating state is overlapped with the first conductive region positioned thereunder through the insulating film, it is possible to realize the effect of the desired object if an electrical connection is realized between the two electrically independent conductive regions, by a resistance with a sufficient degree for discharging charges that typically are a cause of charge-up. FIGS. 2A-2B show section of a hole 62 formed by scanning/irradiating a focused ion beam. In the case where the recess formed as in FIG. 2A is shallow in depth (where the hole is opened shallower in depth than a broadest diameter), the substance sputter-etched is discharged from the hole by irradiating focused ion beam ions 71. However, where the recess formed as in FIG. 2B is deep in depth (where the hole is opened deeper in depth than a broadest diameter) due to the proceeding of sputter etch forming, the formed portion by the focused ion beam is at a deep portion in the hole 62. Accordingly, the discharge speed of the etched substance cannot catch up with the etch speed, and the difference deposited substance 72 adheres to the section thereby making it impossible to form the hole to a depth proportional to forming time. Furthermore, as shown in FIG. 2C, the re-deposited substance 72 left at a deep portion of the hole 62 is stirred by the ion beam into a conductive substance mixed state. As a result, a conductive region can be formed at a side surface. That is, the first conductive layer and the second conductive layer are placed in conduction. When section observation is carried out by a charged particle beam, electric charges are discharged through a newly formed electric path. As a result, charge-up does not occur and a clear interconnect section can be observed. |
claims | 1. A nuclear reactor, comprising:a vessel closed at a top thereof by a radially external fixed closing structure and by a radially internal mobile closing structure, wherein:the vessel contains a core that is immersed in a primary cooling fluid of the vessel; andthe core includes a plurality of fuel elements, control rods, shutdown rods, and a hydraulic separation structure delimiting a hot manifold and a cold manifold;the primary cooling fluid of the core circulates in the hot manifold and the cold manifold and, during normal operation of the nuclear reactor, a first free surface level of the primary fluid in the hot manifold differs from a second free surface level of the primary fluid in the cold manifold;circulation pumps for circulating the primary fluid; andheat exchangers;wherein the control rods and the shutdown rods are inserted in respective penetrations of the fixed closing structure and located radially external to the radially internal mobile closing structure and external to an upper portion of the hydraulic separation structure containing respective heads of the plurality of fuel elements;wherein the control rods and the shutdown rods are inserted in a radially wider lower part of the hydraulic separation structure via respective ducts that extend from a wall of the hydraulic separation structure above the second free surface of the primary fluid in the cold manifold. 2. The nuclear reactor of claim 1, wherein the control rods and shutdown rods have no mechanical connections with the core. 3. The nuclear reactor of claim 1, wherein rotation of the control rods around an axis of a control mechanism modifies a distance of respective neutron absorbers from an active part of the core. 4. The nuclear reactor of claim 1, wherein the shutdown rods are operated by a float which, via a down-up movement, positions respective neutron absorbers near an active part of the core when the circulation pumps of the primary fluid slow down. 5. The nuclear reactor of claim 4, further comprising a non-return device that allows upward displacement of the shutdown rods and blocks downward displacements to prevent falling in an event of untimely acceleration of the circulation pumps. 6. The nuclear reactor of claim 5, further comprising a control mechanism that engages first on a head of the shutdown rod and then on the non-return device allows, by a programmed intervention, safe re-actuation of the shutdown rod and re-location of the respective absorbers in a position farthest from the core. |
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051606968 | summary | Present nuclear waste strategies, centered about geologic repository storage, require geologic stability and separation of wastes from human contact for tens of thousands of years. Transmutation offers the potential for transforming the time scales associated with such storage to hundreds of years or less. Transmutation of long-lived nuclear wastes to short-lived or stable isotopes has been studied for many years. A sampling of illustrative techniques was presented in a recent symposium in a presentation entitled "A Conceptual Study Of Actinide Transmutation System With Proton Accelerator(1) Target Neutronics Calculation," by H. Takada, I. Kanno, T. Takizuka, T. Ogawa, T. Nishida, and Y. Kaneko, Proceedings Of The 2nd International Symposium On Advanced Nuclear Energy Research-Evolution By Accelerators, January 24-26, 1990, Mito, Ibaraki, Japan. The authors describe a transmutation apparatus using keV neutrons which requires large material inventories to achieve significant transmutation rates since cross sections for neutron capture are small at these neutron energies. Moreover, the proton beam is admitted to the subcritical reactor target using a window, which limits the neutron flux available for the process. The direct interaction between the proton beam and the sodium coolant will produce substantial quantities of oxygen, carbon, nitrogen, and hydrogen spallation products, which may combine to generate tar. Finally, degradation of the cladding material for the nuclear waste as a result of proton bombardment may present a lifetime problem. In "Accelerator Molten-Salt Breeding And Thorium Fuel Cycle," by Kazuo Furukawa, Yasuaki Nakahara, Yoshio Kato, Hideo Ohno, and Kohshi Mitachi, Proceedings Of The 2nd International Symposium On Advanced Nuclear Energy Research-Evolution By Accelerators, January 24-26, 1990, Mito, Ibaraki, Japan, the authors describe a windowless apparatus accepting high proton beam currents having GeV energies which are caused to impinge directly on the target materials as in the Takada et al. reference except cooled by molten salt. Transmutation is achieved using keV neutrons where the low cross sections of the neutrons require large inventories to achieve useful transmutation throughput. Additionally, since the thorium is mixed with lithium fluoride, proton spallation will again produce bothersome tars. In "Status Report Of The SIN Neutron Source," by F. Atchison and W. E. Fischer, Proceedings Of International Collaboration On Advanced Neutron Sources (ICANS-VII), Sep. 13-16, 1983, Atomic Energy Of Canada, Limited, Report AECL-8488, the authors disclose a low-power target for low flux neutron production in Pb-Bi from neutron bombardment with subsequent neutron thermalization using heavy water. Heat is removed from the target by thermal convection, and the low power levels also permit the use of a window between the accelerator vacuum and the target. The proton beam strikes the target from below which has advantages for the thermal convection cooling. Accordingly, it is an object of the present invention to efficiently transmute higher actinides and other nuclear wastes. Another object of my invention is to generate power from fertile materials while transmuting the fission and other waste in order to avoid long-term storage. Yet another object of the present invention is to generate tritium drawing on an external electric power source, and without generation of waste requiring long-term storage. SUMMARY OF THE INVENTION To achieve the foregoing and other objects and in accordance with the purpose of the present invention, as embodied and broadly described herein, the apparatus for generating power from fertile nuclear materials and transmuting wastes therefrom to less radioactive species may include a source of high intensity, high-energy beam of protons, a liquid-metal spallation target having an upwardly facing open surface for producing a high neutron flux upon being impacted by high-energy protons, a substantially gas-tight enclosure surrounding the spallation target, windowless apparatus for directing the beam of protons onto the open surface of the spallation target, a neutron moderator for thermalizing neutrons generated from the spallation target, a container for holding fertile nuclear material located within the neutron moderator and spaced apart from and outside of the spallation target, and a container for holding materials to be transmuted located within the neutron moderator and spaced apart and outside of the spallation target, yet closer thereto than the fertile nuclear material container. In another aspect of the present invention, in accordance with its objects and purposes, the apparatus for transmuting higher actinide waste along with .sup.99 Tc, .sup.129 I, and other fission product waste may include a source of high intensity, high-energy of protons, a liquid-metal spallation target having an upwardly facing open surface for producing a high neutron flux upon being impacted by high-energy protons, a substantially gas-tight enclosure surrounding the spallation target, windowless apparatus for directing the beam of protons onto the open surface of the spallation target, a neutron moderator for thermalizing neutrons generated from the spallation target, and a container for holding the material to be transmuted located within the neutron moderator and spaced apart from and outside of the spallation target. In yet another aspect of the present invention, in accordance with its objects and purposes, the apparatus for simultaneously transmuting higher actinide materials and producing tritium may include: a source of high intensity, high-energy protons, a liquid-metal spallation target having an upwardly facing open surface for producing a high neutron flux upon being impacted by high-energy protons, a substantially gas-tight enclosure surrounding the spallation target, windowless apparatus for directing the beam of protons onto the open surface of the spallation target, a neutron moderator for thermalizing neutrons generated from the spallation target, a container for holding the higher actinide materials located within the neutron moderator and spaced apart from and outside of the spallation target, and a container located within the neutron moderator and spaced apart from the spallation target for holding materials which generate tritium upon interaction with neutrons. Benefits and advantages of the present invention include power production from .sup.238 U, .sup.232 Th, and from burning defense and commercial higher actinide waste without long-term waste management, avoiding long term management of commercial spent fuel by fission product transmutation using .sup.239 Pu from reprocessing tritium production while burning higher actinide waste, .sup.238 U and .sup.232 Th. Another benefit of the present invention includes the avoidance of the major problem of current nuclear reactors which is catastrophic runaway. Since the fission products can be continually removed from the molten salt stream and the inventory of fissile material is small, the probability and consequences of a loss-of-coolant accident also are significantly reduced. The probability of such an accident is greatly reduced since this system operates well below nuclear criticality and requires no control rods. 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. |
description | The present application claims priority to Chinese Patent Application No. CN 201410168041.1 filed on Apr. 24, 2014, the disclosure of which is hereby incorporated by reference in its entirety. The present disclosure relates to uranium contamination treatment, and specifically it is a comprehensive joint technology-based device and method for the treatment of uranium-contaminated soil. With China's increasingly rapid exploration and mining of uranium, there emerge numerable uranium tailings and waste ores. Currently, the dedicated storage facility of uranium tailings is the uranium tailings impoundment. According to incomplete statistics, the total amount of uranium tailings worldwide has reached 20 billion tons. Since 1940s, uranium and other radionuclides from a large quantity of in-service or retired uranium tailings have went into the ground water, and flowed and spread with the groundwater, causing serious harm to the ground water and the ecological environment. Uranium is primarily a a-ray source (uranium-235 is 4.679 MeV, uranium-238 is 4.270 MeV), and its half-life period is up to a million years (the half-life of uranium-235 is 704 million years, and the half-life of uranium-238 is 4.46 billion years), with an extremely slow convection velocity. Uranium in the soil will be accumulated in the human body due to the bioaccumulation action, while its chemical behavior and physiological toxicity are similar with lead, which means it will cause chemical toxicity to human and animal kidney. When the volume of uranium in human body is over 3 mg/kg, it will do harm to human body or even cause cancer. The soil contamination caused by uranium is mainly of three kinds. The first kind of contamination is caused by the test and use of depleted uranium bombs. For example, the UK and the US once conducted the depleted uranium bomb tests, respectively leading to 45 t and 70 t depleted uranium substances left in the environment. The second kind of contamination is caused by the production and test of nuclear weapons. The contaminated soil caused by the nuclear weapon production of US Department of Energy is estimated to be 3.0×107-8.0×107 m3, and the contaminated water is of 1.8×109-4.7×109 m3. The third kind of contamination is caused by waste water, waste gas, waste residue, etc. discharged by nuclear power plants. On Mar. 11, 2011, an explosion occurred in Fukushima Daiichi, which used uranium-235 as the fuel for nuclear reactor, and the leaked radioactive material—uranium received much concern all over the world. The fourth kind of contamination is caused by waste residue of uranium tailings. The volume of uranium tailings waste is estimated to reach 9.38×108 m3 due to the uranium mining. China's uranium mines are located in more than 30 counties of 15 provinces and cities nationwide, ⅔ or more of uranium mines are located in mountainous, humid and rainy regions, while nearly ⅓ are located in hills and arid regions. Over the years, the total amount of waste ores mined from uranium mines are about 28 million tons, covering an area of 250 hm3. And uranium tailings charged by waste water treatment plants are about 30 million tons, which is covering an area of 375 hm3 if calculated as its average stacking height of 4 m. Uranium exists in the soil in an adsorbed state (in soil particles and pore water), the compound state, the sediment state and the exchangeable state, while these sates will greatly affect the uranium migration. Due to the huge nuclear pollution hazards, the depollution of nuclear contamination in a large scale is a worldwide problem. All walks of life are looking for ways to clear nuclear contamination, and they have achieved some results. The depollution methods adopted by most countries include: physical methods, chemical methods, electrochemical methods, physical-chemical methods, microbial clearance and soil nuclear contamination removal method, but these methods are usually of a high cost and easy to cause a secondary pollution to the environment, failing to solve the problem of radionuclides removal fundamentally. Although the forest restoration method can remedy the contaminated soil in a large scale, and can be used for the development of industrial raw material forest and firewood forest, but its required radionuclide concentration can only be at a not high level, greatly restricting its application. Foreign scholars mainly adopt chemical method for the treatment of uranium-contaminated soil, for example, Caroline et. al (1998) uses sodium bicarbonate for the extraction of uranium in soil solution, and then adopts the ion exchange technology to remove the uranium in the soil solution. Kantar et al. (2006) points out that the citric acid elution technique can be used to improve the clearance efficiency of uranium in soil under the weak acid condition. And Crean et. al (2013) uses multi-level chemical extraction techniques for the remediation of depleted uranium particles in contaminated soil. Currently, in China, researches on the uranium-contaminated soil are mainly focused on bioremediation, for example, the patent CN200810030860.4 discloses a method of using American phyotalacca acinosa seedling for the treatment of the soil contaminated by uranium tailings; patent CN201210056468.3 discloses a method of using water spinach for the treatment of soil contaminated by high-concentration uranium, and the patent CN201210155561.X discloses a method of using fertilizer to promote the treatment of soil contaminated by uranium or cadmium with broad beans. Considering the current technology deficiency, in this paper a comprehensive joint technology-based device for the treatment of uranium-contaminated soil is provided. In order to achieve the goal above, this invention is adopting the following technical program: The comprehensive joint technology-based device for the treatment of uranium-contaminated soil is of the following structure: Set the cathode zone 5 and anode zone 6 at both sides of uranium-contaminated soil, and set the cathode and anode respectively in those two zones, which are respectively connected to the power transformer 9 with wires; set several sprinklers 10 between cathode zone 5 and anode zone 6 in the form of an array; set pumping well 1 in the vicinity of the cathode zone 5, with pumping pipeline 3 in it, and pumping pipeline 3, pump 11 and pumping controller 12 should be connected to the water collecting sump 13 sequentially; set the soil storage tank 14 near the water collecting sump 13, and connect one end of the remediation-separation system 15 to the soil storage tank 14 and water collecting sump 13, while connect the other end to the soil recovery tank 16, ground water recovery tank 17 and uranium recovery tank 18; set recharge well 2 in the vicinity of the anode zone 6, with the recharge pipeline 4 in it, and the recharge pipeline 4, recharge pump 19 should be connected to the groundwater storage tank 17 sequentially. Said comprehensive joint technology-based device for the treatment of uranium-contaminated soil further comprises the catholyte processing system 7 and anolyte processing system 8. The catholyte processing system 7 comprises a catholyte storage compartment, a catholyte pH controller and a peristaltic pump; the catholyte pH controller is connected to the catholyte storage compartment, while the catholyte storage compartment is connected to the cathode zone 5 through the corresponding pipeline and the peristaltic pump. The anolyte processing system 8 comprises an anolyte storage compartment, an anolyte pH controller and a peristaltic pump; the anolyte pH controller is connected to the anolyte storage compartment, while the anolyte storage compartment is connected to the anode zone 6 through the corresponding pipeline and the peristaltic pump. The remediation-separation system 15 mentioned above is composed by the photoelectric remediation device 22, the screen 23, the solid phase extraction device 24, the aqueous phase processing device 25, and the solid phase processing device 26. One end of the photoelectric remediation device 22 is connected to the water collecting sump 13 and the soil storage tank 14, and the other end is connected to the screen 23; the screen 23 is respectively connected to the aqueous phase processing device 25 and the solid phase extraction device 24; the aqueous phase processing device 25 is connected to the ground water recovery tank 17; the solid phase extraction device 24 is connected to the solid phase processing device 26 and the uranium recovery tank 18; the solid phase processing device 26 is connected to the soil recovery tank 16. The pumping pipeline 3 and the recharge pipeline 4 mentioned above are both non-sand concrete tube wells. Their inner diameter is of 250-500 mm and depth of 35-50 m, and their periphery is filled with backfill and gravels, with the filling layer of 60-100 mm in thickness. The cathode and anode mentioned above are both plate electrodes. A uranium-contaminated soil treatment method with the use of the comprehensive joint technology-based device for the treatment of uranium-contaminated soil mentioned above is of the following steps: Step 1: Install the comprehensive joint technology-based device for the treatment of uranium-contaminated soil at the site contaminated by uranium, with the cathode, anode, pumping pipeline 3 and recharge pipeline 4 inserted into the soil, and the sprinklers 10 above the contaminated soil; Step 2: Spray the eluting solution via multiple sprinklers 10 to the soil until the concentration of eluting solution in the soil is 10-30 ml/g; Step 3: Add the catholyte and anolyte respectively in the cathode zone 5 and the anode zone 6, and maintain the voltage between the cathode and anode of 150-600V via the power transformer 9; keep the electrokinetic remediation for 2-8 days, and then uranium substances will migrate to the vicinity of anode, leading to the gathering of uranium pollutants; Step 4: During the electrokinetic remediation, start up the pump 11 to transport the contaminated ground water in the phreatic layer to the water collecting sump 13 through the pumping pipeline 3, with the pumping volume of 5-10 m3/h and lasting for 1.5-6 days; Step 5: When the electrokinetic remediation is finished, excavating the soil in the area within the diameter of 1-2.5 m and the depth of 10-20 m centered on the anode zone 6, and then transport it to the soil storage tank 14; Step 6: Transport the contaminated soil in the soil storage tank 14 and the contaminated water in the water collecting sump 13 to the remediation-separation system 15, and mix the soil and ground water to form a suspension; adjust the pH of the suspension to 2.8-3.5, and irradiate it with the fluorescent light for 24-36 h for the photolysis remediation, then the uranium compound in the suspension will become the uranyl hydroxide and deposit; after the remediation, deposit and separate the soil, water and uranyl hydroxide, and respectively transport them to the soil recovery tank 16, the ground water recovery tank 17 and the uranium recovery tank 18; Step 7: Transport the soil remedied through the photolysis in the soil recovery tank 16 to the vicinity of the anode zone 6, and start up the recharge pump 19 to transport the remedied ground water in the ground water recovery tank 17 to the recharge well 2 via the recharge pipeline 4; Step 8: Test the soil remediation effect, and repeat Step 1 to Step 7 if needed until the uranium in the soil meets the requirement of the safety standards or is totally removed. Specifically: Before Step 1, conduct the ground investigation of the soil contaminated by the radioactive uranium that remains to be treated, so as to get information of the contaminated soil area, hydrogeological conditions, types of pollutants, and data of the physical and chemical nature, thus determining the contamination status. After Step 1, the pumping pipeline 3 and the recharge pipeline 4 should be 0.5-2 m in depth across the phreatic layer. After Step 1, the cathode and anode should be inserted 10-20 m in depth into the contaminated soil surface, tangent with and the phreatic layer. The used eluting solution in Step 2 is the 0.2-0.6M citric acid solution or sodium bicarbonate solution. The spray duration in Step 2 is 2-4 h. The weight/volume of soil and ground water in the suspension used in Step 6 is 1 g: (10-20) ml. The separation of soil, water and uranyl hydroxide in Step 6 can be achieved as follows: when the photolysis remediation is finished, solid substances and the solution in the suspension will be separated through the screen 23, and the screen whose bore diameter is 0.22 micron is preferred; transport the obtained ground water through the separation to the aqueous phase processing device 25 to adjust a neural pH, and then transport it to the ground water recovery tank 17; transport the obtained soil to the solid phase extraction device 24 to recover the uranyl hydroxide precipitate in the soil with the use of the TBP-sulfonated kerosene extraction technique, and then transport the recovered uranium substances to the uranium recovery tank 18; transport the rest soil to the solid phase processing device 26 to adjust a neural pH, and then transport it to the soil recovery tank 16. In Step 8, adopt the inductively coupled plasma atomic emission spectrometry (ICP-AES) to measure the residual volume of uranium in soil and in the uranium recovered solution, so as to get the soil remediation effect. Beneficial effects of the invention include: (1) This comprehensive joint technology can not only reduce the uranium volume in the contaminated soil or water, but also recycle the obtained uranium; (2) the electrokinetic technique can gather the uranium to the vicinity of the anode, which is convenient for the centralized processing; (3) add the extracting agent, citric acid or sodium bicarbonate solution to the soil can effectively extract the uranium from the soil phase to the aqueous phase; (4) the photolysis can deposit the uranium, which is convenient for further recycling; (5) the treatment process is clean and will not cause secondary pollution with a shorter processing cycle and guaranteed security. Numerals in the drawing represent: 1—pumping well; 2—recharge well; 3—pumping pipeline; 4—recharge pipeline; 5—cathode zone; 6—anode zone; 7—catholyte processing system; 8—anolyte processing system; 9—the power transformer; 10—sprinklers; 11—pump; 12—pumping controller; 13—water collecting sump; 14—soil storage tank; 15—remediation-separation system; 16—soil recovery tank; 17—ground water recovery tank; 18—uranium recovery tank; 19—recharge pump; 20—soil excavating zone; 21—soil surface; 22—photoelectric remediation device; 23—screen; 24—solid phase extraction device; 25—aqueous phase processing device; 26—solid phase processing device. Here, a further introduction of this invention will be made combined with the attached figures and specific examples: FIG. 1 is the diagram of the invented comprehensive joint technology-based device for the treatment of uranium-contaminated soil, which is composed by the pumping system, the electrokinetic remediation system, elution system, remediation-separation system and recharge system. The specific structure is as follows: set the cathode zone 5 and anode zone 6 at each side of the uranium-contaminated soil, and respectively set the cathode and anode inside the cathode zone 5 and anode zone 6, which are respectively connected to the power transformer 9. Control the volume, pH and other parameters of catholyte and anolyte in the cathode zone 5 and anode zone 6 through the catholyte processing system 7 and anolyte processing system 8. The catholyte processing system 7 is composed by a catholyte storage compartment, a catholyte pH controller and a peristaltic pump; the catholyte pH controller is connected to the catholyte storage compartment, while the catholyte storage compartment is connected to the cathode zone 5 through the peristaltic pump. The anolyte processing system 8 is composed by an anolyte storage compartment, an anolyte pH controller and a peristaltic pump; the anolyte pH controller is connected to the anolyte storage compartment, while the anolyte storage compartment is connected to the anode zone 6 through the peristaltic pump. Multiple sprinklers 10 will be set between the cathode zone 5 and anode zone 6 in the form of an array, and when the device is installed at the site of the contaminated soil, these sprinklers can spray on the contaminated soil between the cathode zone 5 and anode zone 6. Set the pumping well 1 in the vicinity of the cathode zone 5, and set the pumping pipeline 3 in it. The pumping pipeline 3 is connected to the pump 11, and the pump 11 and the pumping controller 12 are sequentially connected to the water collecting sump 13, so as to transport the contaminated ground water in the in shallow water layer to the water collecting sump 13; set the soil storage tank 14 in the vicinity of the water collecting sump 13, which is used for the storage of soil after the electrokinetic remediation. One side of the remediation-separation system 15 is connected to the water collecting sump 13 and the soil storage tank 14, while the other side is respectively connected to the soil recovery tank 16, the ground water recovery tank 17 and the uranium recovery tank 18. As is shown in FIG. 2, adopt the remediation-separation system 15 for the photolysis remediation of contaminated ground water and the obtained soil after the electrokinetic remediation, and specifically, the remediation-separation system 15 is composed by the photoelectric remediation device 22, the screen 23, the solid phase extraction device 24, the aqueous phase processing device 25 and the solid phase processing device 26. And one side of the photoelectric remediation device 22 is connected to the water collecting sump 13 and the soil storage tank 14, and the other side is connected to the screen 23, while the screen 23 is respectively connected to the aqueous phase processing device 25 and the solid phase extraction device 24; the aqueous phase processing device 25 is connected to the ground water recovery tank 17; the solid phase extraction device 24 is connected to the solid phase processing device 26 and the uranium recovery tank 18; and the solid phase processing device 26 is connected to the soil recovery tank 16. Set the pumping recharge well 2 in the vicinity of the anode zone 6, and set the recharge pipeline 4 inside it. The recharge pipeline 4 is connected to the recharge pump 19, and the recharge pump 19 is connected to the ground water recovery tank 17. The pumping pipeline 3 and the recharge pipeline 4 adopted by this research are both non-sand concrete tube wells. Their inner diameter is of 250-500 mm and depth of 35-50 m, and their periphery is filled with backfill and gravels, with the filling layer of 60-100 mm in thickness. And the cathode and anode mentioned above are both normal plate electrodes. Remedy the uranium-contaminated soil with the use of said device, and the specific steps are as follows: Step 1: Conduct the ground investigation of the soil contaminated by the radioactive uranium that remains to be treated, so as to get information of the contaminated soil area, hydrogeological conditions, types of pollutants, and data of the physical and chemical nature, thus determining the contamination status. Step 2: Install the comprehensive joint technology-based device for the treatment of uranium-contaminated soil at the site contaminated by uranium, with the cathode, anode, pumping pipeline 3 and recharge pipeline 4 inserted under the surface soil 21, and the sprinklers 10 uranium above the contaminated soil; ensure that the pumping pipeline 3 and the recharge pipeline 4 is 0.95 m in depth across the phreatic layer, and the cathode and anode is inserted 12 m in depth into the contaminated soil surface, tangent with and the phreatic layer. Step 3: Spray the citric acid solution via multiple sprinklers 10 to the contaminated soil, and then measure the volume of eluting solution in different soils (10, 20, 30 ml/g) and the uranium removal efficiency of different eluting duration (2, 3, 4 h). Step 4: Use the catholyte pH controller and the anolyte pH controller to control the pH of catholyte and anolyte, and transport the catholyte and anolyte respectively to the cathode zone 5 and the anode zone 6 with the corresponding pumps, and among them the catholyte and anolyte are preferably graphite electrode solution, purchased through the business approach. Maintain the voltage between the cathode and anode of 150V via the power transformer 9, and keep the electrokinetic remediation for two days so as to make uranium substances migrate to the vicinity of anode, leading to the gathering of uranium pollutants; Step 5: During the electrokinetic remediation, start up the pump 11 to transport the contaminated ground water in the phreatic layer to the water collecting sump 13 through the pumping pipeline 3, with the pumping volume of 5 m3/h and lasting for 1.5 days; Step 6: When the electrokinetic remediation is finished, excavating the soil in the area within the diameter of 1-2.5 m and the depth of 10 m centered on the anode zone 6, just as the soil excavating zone 20 shown in FIG. 1, and then transport it to the soil storage tank 14; Step 7: Transport the contaminated soil in the soil storage tank 14 and the contaminated water in the water collecting sump 13 to the remediation-separation system 15, and mix the soil and ground water to form a suspension, while weight/volume of soil and ground water is 10 g:120 ml. Adjust the pH of the suspension to 3.5, and irradiate it with the fluorescent light for 24 h for the photolysis remediation, then the uranium compound in the suspension will become the uranyl hydroxide and deposit. After the remediation, deposit and separate the suspension with the 0.22 micron screen 23. Transport the obtained ground water to the aqueous phase processing device 25 to adjust a neural pH, and then transport it to the ground water recovery tank 17. Transport the obtained soil to the solid phase extraction device 24 to recover the uranyl hydroxide precipitate in the soil with the use of the TBP-sulfonated kerosene extraction technique, and then transport the recovered uranium substances to the uranium recovery tank 18. Finally, transport the rest soil to the solid phase processing device 26 to adjust a neural pH, and then transport it to the soil recovery tank 16. Step 8: Transport the soil remedied through the photolysis in the soil recovery tank 16 to the vicinity of the anode zone 6, and start up the recharge pump 19 to transport the remedied ground water in the ground water recovery tank 17 to the recharge well 2 via the recharge pipeline 4; Step 9: Adopt the inductively coupled plasma atomic emission spectrometry (ICP-AES) to measure the residual volume of uranium in soil and in the uranium recovered solution, so as to get the soil remediation effect. Repeat Step 2 to Step 8 if needed until the uranium in the soil meets the requirement of the safety standards or is totally removed. The test result is shown in the following table: TABLE 1The uranium removal efficiency of differenteluting duration and volumesEluting solution volume in soil (ml/g)102030Eluting duration (h)234234234Uranium51.36260.963.475.183.769.780.279.8removalefficiency(%) Specific approaches are the same with the implementation example 1, but change the voltage and remediation duration of the electrokinetic remediation, and then analyze the uranium removal efficiency changes of ICP-AES under the different voltages (200V, 300V, 400V, 500V, 600V) and electrokinetic remediation duration (2, 4, 6, 8 day). The result is shown in FIG. 3. It can be seen that increasing the voltage or prolonging the electrokinetic remediation duration time will help to improve the removal efficiency of uranium in soil, and when the voltage is 500V with the electrokinetic remediation time of 8 d, the repair efficiency will be up to 82.9%. |
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abstract | The structures of base semiconductor materials such as Si are modified by the use of isotope transmutation alloying. A radioisotope such as Si31 is added into a base semiconductor material such as Si, and the radioisotope is transformed to a transmuted form within the crystal lattice structure of the base semiconductor material. A master alloy comprising a relatively large amount of radioisotope such as Si31 may initially be made, followed by introduction of the master alloy into the base semiconductor material. When Si31 is used as the radioisotope, it may be transmuted into P31 within an Si crystal lattice structure. Metastable semiconductor materials doped with otherwise insoluble amounts of selected dopants are produced as a result of the transmutation process. |
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abstract | A portable orthovoltage radiotherapy system is described that is configured to deliver a therapeutic dose of radiation to a target structure in a patient. In some embodiments, inflammatory ocular disorders are treated, specifically macular degeneration. In some embodiments, the ocular structures are placed in a global coordinate system based on ocular imaging. In some embodiments, the ocular structures inside the global coordinate system lead to direction of an automated positioning system that is directed based on the ocular structures within the coordinate system. |
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description | This application is a Section 371 National Stage Application of International Application No. PCT/GB2005/002894, filed 25 Jul. 2005 and published as WO 2006/013325 on 9 Feb. 2006, in English. The present invention relates to X-ray apparatus, in particular X-ray apparatus that has an investigative function. Such X-ray apparatus can be stand-alone X-ray apparatus, for purely investigative purposes, or can be integrated as part of an investigative function provided on a radiotherapeutic apparatus. Computed Tomography scanning is a well known diagnostic technique and, in Its cone beam form, involves directing a wide beam of X-rays towards and through the patient and capturing the resulting two-dimensional image on a flat panel detector behind the patient. The apparatus (source and detector) is then rotated around the patient to obtain a multiplicity of images from different directions. These images are combined via a suitable computing means in order to produce a three-dimensional representation of the internal structure of the patient. One limiting factor is the cost and size of the detector. Flat panel X-ray detectors are typically very expensive, and the cost increases with the dimensions of the detector. In practice this places an upper limit on the possible size of the flat panel detector. This in turn places a limit on the maximum aperture of the apparatus. There are ways to increase the effective aperture of the device, within limits. Normally, the cone beam is directed along a central beam axis that coincides with the isocentre of the device, and the flat panel detector is centred on that beam axis. This will mean that each successive image taken by the flat panel detector will show a section of the patient centred on the isocentre. These can then be reconstructed in the normal way. However, for particularly large patients this aperture may be insufficient. In this case, the aperture of the apparatus can be increased by moving the flat panel detector such that the central beam axis intersects near to one edge of the detector. This X-ray beam can then be collimated differently so that the cone beam is offset from the (previous) central beam axis and still covers the area of the flat panel detector. The beam will then be centred on an offset beam axis. In this case, each individual image will only show half of the relevant portion of the patient. However, after the apparatus has rotated through 180°, the other half will be brought into the image. When these images are reconstructed using a suitably reconstructed algorithm, a complete rendering of the patient will still be possible, albeit with a lower resolution reflecting the fact that each voxel of the reconstructed volume has been reconstructed using only half the amount of data. Whilst this arrangement is potentially beneficial in that it allows a different compromise to be reached between aperture and image quality in cases that demand it, it does place some limitations on the apparatus design. In particular, the X-ray tube must be able to provide a beam that is of twice the width otherwise required. This limits the choice and specification of X-ray tubes that can be used and may impose difficulties in other areas, in that a tube that is able to provide a sufficiently wide beam may be inadequate in other ways. The present invention therefore provides an investigative X-ray apparatus, comprising a source of X-rays emitting a cone beam centred on a beam axis, a collimator to limit the extent of the beam, and a two-dimensional detector, the apparatus being mounted on a support which is rotatable about a rotation axis, the collimator having a first state in which the collimated beam is directed towards the rotation axis and the second state in which the collimated beam is offset from the rotation axis, the two-dimensional detector being movable accordingly, the beam axis being offset from the rotation axis by a lesser amount than the collimated beam in the second state. Thus, in effect, the X-ray source is given a permanent offset of a few degrees (such as 3-4°) such that its natural axis is halfway between the two extremes called for by the collimator. The X-ray source is no longer directed towards the isocentre as would normally be the case. Thus, in an alternative aspect, the present invention provides an investigative X-ray apparatus comprising an X-ray source and a collimator set in which the beam is not orthogonal to the collimators. This is advantageous in that the entire field of the X-ray tube can be utilised. X-ray tubes typically have edge effects such as tube heel, and this can be kept away from both potential images. As a result, a lesser field is required of the X-ray tube and the choice of tube designs and capacities can be widened so as to optimise the performance of the X-ray tube in other aspects. FIG. 1 shows a typical radiotherapy machine. This has a rotatable support 10 on which is mounted a therapeutic X-ray source 12 which is able to produce a collimated beam of high energy X-rays 14 centred on a therapeutic beam axis 16. Also mounted on the rotatable support 10 is an investigative X-ray source 18, which produces a beam of low-energy X-rays 20 along an investigative beam axis 22. On the opposite side of the support 10, a flat panel detector 24 is positioned so as to intersect with the investigative beam axis 22. The rotatable support 10 is arranged to rotate about an axis which passes through the coincidence of the therapeutic beam axis 16 and the investigative beam axis 22, and which is orthogonal to both axes. In this case, the therapeutic beam axis and the investigative beam axis are orthogonal to each other, but this is not essential and other designs are possible. The point of coincidence of the two beam axis 16, 22 and the rotation axis of the support 10 is referred to as the “isocentre”. A patient table 26 is located slightly below the isocentre, and a patient 28 resting on the patient table will therefore just lie at the isocentre of the apparatus. In practice, the patient table 26 is made so as to be moveable, to allow the patient to be positioned relative to the isocentre, and permit the treatment of tumours at a variety of bodily locations. During treatment, the therapeutic X-ray source 12 is activated and the beam 14 is collimated so as to match the shape of the tumour. The rotatable support 10 can be used to rotate the therapeutic X-ray source 12 around the patient so as to direct the beam 14 towards the patient from a variety of directions. Provided that the tumour is at or near the isocentre, it will always be Irradiated. However, the use of a variety of irradiation directions is one factor in reducing the dosage given to healthy tissue whilst maximising the dosage given to the tumour. It is of course essential to ensure that the patient is correctly positioned prior to treatment. To do so, the investigative X-ray source 18 is activated and the low energy beam 20 is passed through the patient and, after attenuation by the patient, is detected by the flat panel detector 24. This produces a two-dimensional projection image of the patient. The rotatable support 10 is then used to rotate the investigative X-ray source 18 and the flat panel detector 24 around the patient thereby producing a collection of projected images showing the patient from every variety of directions. These can be reconstructed using known algorithms to produce a three-dimensional image of the patients adhering structure, the process known as computed tomography or CT scanning. This internal image of the patient can be used as a final check that the patient is in the correct position, and potentially, as a source of feedback to allow fine adjustment of the position of the patient table 26. In FIG. 1, the investigative beam 20 is shown collimated so that the image it projects covers the entire working surface of the flat panel detector 24. As a result, the width of the beam 20 at the patient 28 is large enough to ensure that the whole of the patient 28 is included in the image obtained by the flat panel detector 24. Problems can arise in the case of very large patients, part of whom will lie outside the beam 20. In general, it is not possible simply to select a larger flat panel detector 24 and allow a wider beam, since the flat panel detector 24 is a high value item and larger examples cannot be procured at economic cost. Accordingly, larger patients are dealt with as shown in FIG. 2. The same flat panel detector 24 is moved on its support to an offset position, as shown. Whilst the flat panel detector 24 still coincides with the investigative beam axis 22, that axis 22 now crosses the flat panel detector 24 near one edge of the detector 24. The investigative beam 20 is now collimated slightly differently so that it is no longer centred on the investigative axis 22 but extends from that axis 22 and to one side. As a result, the beam 20 produces an image of approximately one half of the patient 28, in this case the half lying above the isocentre. However, as the apparatus is rotated around the patient 28, after a total rotation of 180° the image will show the area of the patient below the isocentre. FIGS. 3 and 4 illustrate the point schematically. In FIG. 3, the solid vertical line 30 shows the section of the patient which is being viewed at the start of the rotation process. This is represented as a line, whereas the images are of course projected images rather than a section, but FIG. 3 illustrates the principle only. As the apparatus rotates, the effective image moves through an angle to the dotted line 32, and as rotation continues further the images moves to the dotted line 34. Thus, as rotation continues, the image taken of the patient maps out a cylindrical volume centred on the axis of rotation. FIG. 4 illustrates the offset method. A solid line 36 of identical length to the solid line 30 is again rotated, but this time the axis of rotation is at one end of the solid line 36. Thus, as the image rotates through 38 and 40 etc., a larger cylinder is mapped out. This caters for the larger patient. However, it can be seen that twice as many lines are required to map out the same cylindrical volume. Thus, the offset rotation arrangement must either spend twice as long gathering images in order to produce the same quality CT reconstruction, or must accept a lower quality CT reconstruction deriving from fewer images. This choice is however clinically useful. FIG. 5 shows an improved apparatus for use in this type of diagnosis. This comprises, in general, an X-ray generating source 50 and a collimator set 52. The investigative beam axis 22 is shown, together with the isocentre 54 and the flat panel detector 24. The X-ray source 50, which we will described in more detail later, produces a beam 56 which is then collimated in the collimator 52. In this design of collimator set 52, a number of slots 58, 60 are provided to receive collimators and filters as required. The first slot 58 contains a beam collimator 62 to produce the investigative beam 20 from the output beam 56 of the source 50, so that the beam 20 just covers the flat panel detector 24. In this case, as shown, the beam collimator 62 collimates the beam 56 evenly, by reducing its width equally on both sides. As shown in FIG. 5, the collimator 52 is aligned with the beam axis 22 and the isocentre 54. However, the X-ray source 50 is offset by an angle 66 from being perfectly orthogonal to the investigative beam axis 220. As a result, the X-ray source 50 and the collimator 52 are not in alignment, and the approximate centre 68 of the beam does not coincide with the isocentre 54. However, the beam does extend across the beam axis 22 and the isocentre 58 is included within the extent of the beam 20. FIG. 6 shows the same apparatus in which an alternative beam collimator 62′ has been fitted, together with an alternative filter 64′. The second filter 64′ differs only in that its centre is suitably offset. The alternative beam collimator 62′ differs in that it collimates the beam asymmetrically with respect to the beam 56 emanating from the X-ray source 50, but nevertheless symmetrically about the beam axis 22 and the isocentre 54. In this way, the apparatus can be used as described with respect to FIG. 1. FIG. 7 shows the same apparatus with a still further alternative collimator 62″ and filter 64″. In this case, the collimator 62″ collimates the beam 56 asymmetrically, but this time in the opposite sense to that of FIG. 6. Instead of returning the beam towards the isocentre 54, the beam is offset still further from the isocentre 54 such that the beam 20 only just overlaps with the isocentre 54. This produces an offset beam for use in the manner as described with respect to FIG. 2 above. It will be appreciated that the two extremities of collimation that are required in clinical practice, as shown in FIG. 6 and FIG. 7 respectively, now occupy the extremities of the usable area of the beam. The fullest available extent 56 of the beam is therefore used, by virtue of the angle 66 between the X-ray source 50 and the collimator set 52. This relieves the designer of the need to select an X-ray tube on the basis of its wide available field, and allows the optimisation of the X-ray tube based on other requirements of the device. FIG. 8 shows, for information, a typical target 70 for use in an X-ray source 50. FIG. 9 shows the apparatus, schematically, including the target 70. A high voltage source 72, typically providing 150 kV is arranged to produce a potential between a hot filament 74 and the anode 70. The anode 70 is itself mounted on a spindle 76 which is rotatable by a motor 78. Thus, a beam of electrons 80 travels from the filament 74 towards the anode 70. The anode 70 has a molybdenum core 82 with a generally circular face on which is mounted an annular ring 84 of tungsten/rhenium target material. The apparatus is disposed such that the electron beam 80 lands on the anode at the target material 84. The surface is slightly bevelled so that, in respect of the surface, the electron beam 80 arrives at an angle and, as a result, an emitted beam of X-rays 86 departs the anode 70 in a direction which is roughly perpendicular to the incoming electron beam 80. This beam 86 is then collimated by suitable beam stops 88 to produce the output beam 56 of the X-ray source. The motor 78 drives the anode 70 via the spindle 76 so that the annular target 84 is constantly rotating. As a result, the point of contact of the incoming electron beam 80 is constantly moving across the anode although the rotationally symmetric design of the anode 70 means that this does not affect the output beam 56. As a result, the anode 70 is better able to cool notwithstanding the energy absorbed from the electron beam 80. The entire apparatus of FIG. 9 is typically enclosed within a suitable vacuum flask, which is itself suspended in a bath of flowing oil so as to assist in heat removal. It will be appreciate from FIGS. 8 and 9 that it is only possible to widen the output beam 56 within limits. The width of the output beam 56 will in practice be limited by the size of the rotating anode 70 and by the geometry of the apparatus, for example of the direction of the electron beam 80, the degree to which the target surface 84 is bevelled, and the dimensions of the target surface 84. Limitations such as the need to rotate the anode 70 and the requirement that the anode be adequately cooled mean that there are limits to the available width of the beam 56. Beyond those available limits, the X-ray Intensity becomes less uniform as it eventually fades away to nothing, a phenomenon known as “tube heel”. Accordingly, the invention as described with respect to FIGS. 5, 6 and 7 allows the available extent of the X-ray beam to be used more efficiently, thereby relaxing the design requirements placed on this aspect of the X-ray tube and allowing it to be optimised in other respects. It will of course be understood that many variations may be made to the above-described embodiment without departing from the scope of the present invention. |
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050911422 | description | DESCRIPTION OF PREFERRED EMBODIMENT FIG. 1 shows a fuel assembly 1 consisting of a bundle of parallel fuel rods 2 held by struts 3 arranged with certain spacing along the length of the rods 2. The struts 3 consist of grids, the cells of which each receive a fuel rod. Certain positions in the latticework of the grids are occupied by guide tubes 4, which are longer than the fuel rods 2. The guide tubes 4 are connected at one of their ends to an end block 5 forming the top end block of the fuel assembly and at their other end to a second end block 6 forming the bottom end block. When the assembly is in the underwater storage position inside a well, the top end block 5 is accessible from the top of the well. This top end block 5 carries leaf springs 7 ensuring holding of the assembly inside the reactor core, the upper core plate of which rests on the springs 7. The end block also comprises studs 8 projecting relative to its upper surface. FIG. 2 shows the framework 9 of the fuel assembly comprising guide tubes 4, struts 3 and the end blocks 5 and 6. This framework 9 serves as a housing for the fuel rods 2 of the bundle which may be introduced or extracted from the framework when the top end block 5 is removed. In order to effect replacement or removal of rods, demountable connections between the end of the guide tubes 4 and the top end block 5 are provided. FIGS. 3 and 4 show a top end block of a fuel assembly comprising an adapter plate 10 into which the guide tubes engage, inside openings 11 passing through this adapter plate and accessible from the top part of the fuel assembly. The instrumentation (side tube of the fuel assembly situated in the central part is received inside an opening 12 having a special shape. The top end block of the assembly consists of the adapter plate 10 and a frame 14 connected together by means of a skirt 13 welded onto the plate 10 and onto the frame 14. The frame 14 has bosses 8 comprising centering openings and flanges 8' for holding the springs 7. As can be seen in FIG. 4, the through-holes 11 allowing fixing of the guide tubes 4 are arranged in defined positions corresponding t the positions of the twenty-four guide tubes of the assembly. Water flow holes 15 pass through the adapter plate 10 of the end block between the through-openings 11 of the guide tubes. FIG. 5 shows a locking sleeve 20, facilitating implementation of the method according to the invention. This locking sleeve comprises a lower part 24 having an external surface of frustoconical shape which forms the part of the locking sleeve ensuring expansion of the guide tube and an upper part consisting of a cylindrical ferrule 25 forming the ferrule for fixing the locking sleeve in the end block of the fuel assembly. The ferrule 25 consists of six cylindrical segments 25a, 25b, 25c, 25d, 25e and 25f separated from one another by slits 26 arranged in the direction of the generatrices of the ferrule 25. The cylindrical segments 25a to 25f are identical, the slits 26 being arranged at 60.degree. from one another about the axis of the locking sleeve 20. The expansion bush 24 of the locking sleeve, which is frustoconical in shape, comprises an internal bore of substantially cylindrical shape, such that the thickness of the wall of this bush 24 decreases from its large base at the level of which the ferrule 25 is connected, along the line 30, to its small base, at the level of which an inclined frustoconical surface 24a is provided, facilitating engagement of the locking sleeve inside the guide tube. The expansion bush 24 of the locking sleeve, which is relatively thick, has sufficient rigidity to ensure effective holding of the guide tube, as will be explained herein below. The ferrule 25 is substantially less thick than the expansion bush 24, in particular in the vicinity of the junction line 30, i.e., the large base of the part 24. The locking sleeve 20 has an internal radially projecting shoulder 28 ending in an inclined frustoconical surface 28a directed towards the inside of the sleeve. The sleeve 20 may also be made as a single machined piece or, alternatively, the ferrule 25 may be mounted on the top part of the frustoconical holding part 24. The slits 26 may be cut from the ferrule 25, after machining or assembly of the sleeve. In all cases, the cylindrical segments 25a to 25f forming the ferrule 25 may be deformed by pushing inside locking cavities in the end block of the fuel assembly, and by folding about a line situated in the vicinity of the line 30. FIG. 6 shows the adapter plate 10 of a top end block of a fuel assembly, in the vicinity of a through-opening 11 in which the end 4a of a guide tube 4 is engaged. The upper end 4a of the guide tube 4 has been formed prior to introduction into the opening 11, so as to fit perfectly in the inlet part of this opening which has a slightly frustoconical shape and which comprises an annular enlargement 16. The upper part 4a of the tube 4 has a frustoconical shape corresponding to the shape of the inlet part of the opening 11, and an annular securing part 17, radially projecting outwards, intended to fit into the annular enlargement 16 of the plate 10, when the tube 4 is positioned inside the opening 11. The frustoconical inlet part of the opening 11 ends in a shoulder 21 projecting radially towards the inside of the opening 11. It should be noted that the end of the tube 4, when the latter is completely engaged inside the opening 11, does not rest on the shoulder 21. As a result of preforming, the tube car: be arranged in place without difficulty and very precisely inside the end block without the formation of any tearing support. The opening 11 has, at its inlet end on the lower surface of the adapter plate 10, a frustoconical enlargement 19 for facilitating introduction of the tube 4. The upper part or outlet part of the opening 11, situated above the shoulder 21, comprises a ring-shaped radial cavity 22. Above the cavity 22, the opening 11 forms an outlet part 23 widening out and emerging on the upper surface of the adapter plate 10. The reassembly of a fuel assembly comprising guide tubes and connection means as shown in FIGS. 6 and 7 may be performed underwater in the storage well of the nuclear reactor fuel assemblies, as follows. The fuel assembly rests via its bottom end block on the bottom of the well, the guide tubes of the assembly being in the vertical position. The top part of these guide tubes onto which the adapter plate 10 of the end block 5 is engaged is located at a water depth sufficient to protect the operators carrying out the reassembly. Engagement of the adapter plate onto the ends 4a of the guide tubes does not pose any difficulty, the guide tubes being held in a precise transverse arrangement by the struts of the assembly. Moreover, as a result of the preformed ends of the guide tubes, the guide tubes can be engaged and arranged in position precisely inside the through openings of the adapter plate. The top part 4a of each of the guide tubes is made so as to be deformable radially in order to allow it to be engaged inside the opening 11 and then expanded so as to secure the projecting parts 17 inside the annular enlargements 16. To this end, at least two slits such as 18 arranged at 180.degree. relative to one another separate the top part 4a of the guide tube into at least two frustoconical segments deformable in the radial direction. After arranging in position the top end block on the end of the guide tubes, these guide tubes are fixed by means of locking sleeves 20, such as those shown in FIG. 5 and using a tool 31 such as that shown in FIGS. 6 and 7. The tool 31 consists of a pole of considerable length, the bottom part 32 of which can be seen in FIGS. 6 and 7. The pole of the tool 31 is fixed at the top to a holding and raising means such as a rolling bridge serving the fuel well, and may be displaced inside well, with its axis ZZ' in the vertical position. The bottom part 32 of the pole comprises, in sequence from top to bottom, a cylindrical part, a frustoconical bearing part 33 widening outwards, a frustoconical thrusting surface 34 directed inward, and a cylindrical end part 35. The end part 35 of the bottom part 32 of the pole comprises means for retaining a locking sleeve (not shown). These retaining means may consist, for example, of curved leaf springs, the convexity of which is directed outwards and which are fixed in the axial direction, on the external surface of the end part 35. Such retaining devices are described in a copending patent application filed by applicant on the same day as the present application. The tool 31 comprises two tubular-shaped components (37) and 38, the first of which 37 is slidably mounted on the cylindrical part of the pole situated above the frustoconical surface 33, and the second of which (38) is slidably mounted around the first. The first sliding component 37 comprises slits in its bottom part defining radially deformable segments 39, the bottom part 40 of which forms a gripping jaw. The component 37 on which the component 38 is slidably mounted forms a clamping chuck, the bottom parts 39 of which may be displaced radially between a first position shown in FIG. 6, and a second position shown in FIG. 7. The sliding jacket 38 is connected, at the top (not shown), to an axial displacement device allowing it to be displaced between its respective positions shown in FIGS. 6 and 7, respectively. Each of the gripping jaws 40 of the component 37 comprises a frustoconical internal bearing surface 40a and a frustoconical external bearing surface 40b. The sliding jacket 38 comprises a frustoconical bearing surface 38a at the bottom. In order to carry out reassembly of the end block of a fuel assembly, use is made of a sleeve supplying device arranged inside the well, in the vicinity of the end block to be reassembled. Such a supply device has been described in a copending patent application filed by applicant on the same day as the present application. The tool 31 is positioned so as to be able to introduce the end part 35 of the pole into the cylindrical internal bore of the holding bus 24 of a sleeve located in the supply device. The end part 35 engaged inside the sleeve, as a result of its retaining devices, enables the sleeve to be removed and introduced, via the widened out end 23, into an opening 11 of the adapter plate 10 inside which a guide tube 4 is engaged. A thrusting force is exerted on the pole of the tool 31, the bottom part 32 of which is in contact with the shoulder 28 and the surface 28a of the sleeve 20, by means of a shoulder and the frustoconical bearing surface 34, respectively. The thrusting force transmitted to the sleeve 20 enables the latter to be introduced inside the tube 4 in the locked position, as shown in FIG. 6. An adjusting wedge enables the position of the pole of the tool 31 to be precisely determined in order to introduce the sleeve 20. The pole of the tool 31 is then raised so as to disengage its end part 35 from the sleeve 20, which remains in position inside the guide tube which it fixes. Using a known type of tool, a part of each of the cylindrical segments 25a to 25f is pushed into the cavity 22 of the top part of the opening 11 of the plate 10 of the end block. The deformed parts 29 thus obtained, engaged inside the cavity 22, enable the sleeve 20 to be axially locked relative to the end block. The guide tube 4 is thus perfectly fixed inside the demountable end block. In order to carry out demounting of the end block, the locking sleeves 20 engaged inside the guide tubes and fixed inside the adapter plate of the end block must be extracted. For this purpose, a tool 31 is used, which initially is brought underneath the assembly located underwater inside the well and engaged inside the sleeve 20, as shown in FIG. 6. The pole of the tool 31 abuts against the shoulder 28 and the bearing surface 28a via the shoulder of the part 33 and the bearing surface 34. The gripping jaws 40 of the chuck member 37 are then engaged inside the frustoconical opening 23, around the top end of the ferrule 5. The actuating jacket 38 is lowered, thereby radially folding the segments 39 of the component 37 inwards. FIG. 7 shows the final position of the segments 39, at the end of the downward movement of the jacket 38. During their pivoting movement inwards, the segments 39 come into contact with the top part of the cylindrical segments 25a to 25f of the ferrule 25, which they fold inwards until the segments come into contact with the frustoconical surface 33. Towards the end of its downward movement, the actuating jacket 38 comes into contact, via its surface 38a, with the bearing surface 40b of the component 37, which has a certain freedom of sliding movement around the end 32 of the pole. As a result of the downward sliding movement of the component 37, folding of the segments of the ferrule 25 is facilitated and the ferrule 25 is properly gripped by the gripping jaws 40, as shown in FIG. 7. Folding of the segments of the ferrule 25 has allowed the deformed parts 29 of these segments to be disengaged from the cavity 22. The locking sleeve 25 is therefore no longer held inside the opening of the adapter plate by the fixing ferrule. The sleeve 20 may be easily extracted by raising the pole of the tool 31, using the raising means fixed to the top part thereof. It is sufficient, in fact, to overcome the clamping forces of the frustoconical holding bush 24 inside the tube 4, these forces being very much less than those required for disengaging the deformed parts such as 29 from the radial cavity. Moreover, the ferrule 25 is firmly held by gripping between the bearing surfaces 40a of the gripping jaws 40 and the frustoconical surface 33. Extraction of the locking sleeves is therefore obtained without difficulty as a result of the method of the invention. It is possible to use the method according to the invention to perform the extraction of locking sleeves of a type different from that which has been described in which the fixing ferrule consists of successive cylindrical segments separated by slits. The method according to the invention can also be used in the case of sleeves comprising deformable tongues, cut in the fixing ferrule, which may be pushed outwards into cavities of the end block when the locking sleeve is fixed, and folded inwards in order to disengage them from the cavities, when the sleeve is extracted. It is also possible to use a fixing ferrule having folding lines allowing the parts of the ferrule between two folding lines to be folded down and the deformed parts to be disengaged from the corresponding cavities. In the case where the fixing ferrule consists of cylindrical segments, separated by slits, there may be any number of segments, and only part of these segments may comprise parts deformed inside cavities provided in the end block. The tool for performing extraction of the locking sleeves may be different from that which has been described. This tool may carry out only extraction of the sleeve, or both positioning and extraction. The tool may also consist of a plate carrying a set of extraction spindles arranged in a latticework reproducing the arrangement of the guide tubes of one or more fuel assemblies and each formed in the same manner as the extraction tool described above. The plate is associated with means for displacement in the vertical direction and comprises means for simultaneous operation of the folding and gripping components of the set of spindles. The tool is arranged in position on the top end block of at least one fuel assembly, such that the spindles are each engaged onto a sleeve for locking a guide tube of the assembly. The disengagement and extraction of the locking sleeves from the guide tubes of one or more fuel assemblies may therefore be performed simultaneously, thereby ensuring a considerable saving in time. Finally, the method according to the invention applies to any demountable fuel assembly for a lightwater nuclear reactor, in which the guide tubes are fixed inside an end block by means of locking sleeves comprising a ferrule for fixing the sleeve inside the end block. |
053032730 | claims | 1. An apparatus for assembling a nuclear fuel assembly which includes a grid for supporting a plurality of fuel rods, said grid having a plurality of straps intersecting each other to define a plurality of grid cells therein, and a plurality of pairs of dimples and springs provided on the straps for supporting the fuel rods, the dimple and the spring being disposed in facing relation to each other, on wall portions of the straps, which cooperate with each other to define the grid cells, the dimple and the spring projecting into the grid cell, said apparatus comprising: deflecting means disposed adjacent to said grid for deflecting the spring away from the dimple opposing thereto; said deflecting means including a tubular member defining a plurality of circumferentially divided sleeve pieces, a rod member releasably inserted in said tubular member for sliding movement therealong, and drive means drivingly connected to said rod member for moving said rod member in said tubular member in a longitudinal direction thereof to bring the rod member into urging engagement with said sleeve pieces of the tubular member, whereby the sleeve pieces are deflected to be urged against the spring to deflect the same; said assembling apparatus further comprising: holding means for holding said grid; maintaining means disposed adjacent to said holding means for maintaining the spring deflected, said maintaining means including a key member adapted to be releasably inserted in said grid cell of said grid for maintaining the spring deflected, and inserting means for inserting said key member into said grid cell of said grid; and moving means attached to said holding means for moving said holding means into a shifted position, to thereby allow said maintaining means to insert another key member in another grid cell of said grid to maintain another spring deflected. deflecting means disposed adjacent to said grid for deflecting the spring away from the dimple opposing thereto, said deflecting means including a tubular member defining a plurality of circumferentially divided sleeve pieces, a rod member releasably inserted in said tubular member for sliding movement therealong, and drive means drivingly connected to said rod member for moving said rod member in said tubular member in a longitudinal direction thereof to bring the rod member into urging engagement with said sleeve pieces of the tubular member, whereby the sleeve pieces are deflected to be urged against the spring to deflect the same; wherein said deflecting means further includes a first support member carrying said rod member and being movable toward and away from said grid, and a second support member carrying said tubular member and being movable toward and away from said first support member, said first support member and said second support member being operably associated with each other. deflecting means disposed adjacent to said grid for deflecting the spring away from the dimple opposing thereto, said deflecting means including a tubular member defining a plurality of circumferentially divided sleeve pieces, and a rod member releasably inserted in said tubular member for sliding movement therealong; first and second parallelly positioned support members, said rod member being attached to said first support member, said second support member having an aperture aligned with said rod member so as to permit the insertion of said rod member through said aperture, said tubular member being attached to said second support member and being aligned with said aperture for permitting said rod member inserted through said aperture to be inserted into said tubular member for permitting said sliding movement; and drive means connected to said first support member for moving said rod member in said tubular member in a longitudinal direction thereof to bring the rod member into urging engagement with said sleeve pieces of the tubular member, whereby the sleeve pieces are deflected to be urged against the spring to deflect the same. 2. An assembling apparatus as recited in claim 1, wherein said inserting means includes supplying means for supplying the key member into an insertion position, and transfer means for moving the key member arranged in the insertion position into the grid cell. 3. An assembling apparatus as recited in claim 1, wherein said deflecting means further includes a first support member carrying said rod member and being movable toward and away from said grid, and a second support member carrying said tubular member and being movable toward and away from said first support member, said first support member and said second support member being operably associated with each other. 4. An assembling apparatus as recited in claim 3, wherein said drive means of said deflecting means includes a drive mechanism connected to said first support member for moving said first support member in a longitudinal direction of the rod member, and linkage means attached to said first support member and said second support member for associating the movement of said second support member with the movement of said first support member and dissociating the movement of the second support member therefrom. 5. An assembling apparatus as recited in claim 4, wherein said linkage means includes a connecting rod connected at one end to said first support member and at the other end to said second support member, said connecting rod being arranged such that said first support member and said second support member are movable toward and away from each other and that said second support member is movable following the movement of said first support member. 6. An assembling apparatus as recited in claim 5, wherein said linkage means further includes stopper means disposed adjacent to said first and second support members for stopping said second member in association with the movement of said first support member. 7. An apparatus for assembling a nuclear fuel assembly which includes a grid for supporting a plurality of fuel rods, said grid having a plurality of straps intersecting each other to define a plurality of grid cells therein, and a plurality of pairs of dimples and springs provided on the straps for supporting the fuel rods, the dimple and the spring being disposed in facing relation to each other, on wall portions of the straps, which cooperate with each other to define the grid cells, the dimple and the spring projecting into the grid cell, said apparatus comprising: 8. An assembling apparatus as recited in claim 7, wherein said inserting means includes supplying means for supplying the key member into an insertion position, and transfer means for moving the key member arranged in the insertion position into the grid cell. 9. An assembling apparatus as recited in claim 8, wherein said drive means of said deflecting means includes a drive mechanism connected to said first support member for moving said first support member in a longitudinal direction of the rod member, and linkage means attached to said first support member and said second support member for associating the movement of said second support member with the movement of said first support member and dissociating the movement of the second support member therefrom. 10. An assembling apparatus as recited in claim 9, wherein said linkage means includes a connecting rod connected at one end to said first support member and at the other end to said second support member, said connecting rod being arranged such that said first support member and said second support member are movable toward and away from each other and that said second support member is movable following the movement of said first support member. 11. An assembling apparatus as recited in claim 10, wherein said linkage means further includes stopper means disposed adjacent to said first and second support members for stopping said second member in association with the movement of said first support member. 12. An apparatus for assembling a nuclear fuel assembly which includes a gird for supporting a plurality of fuel rods, said grid having a plurality of straps intersecting each other to define a plurality of grid cells therein, and a plurality of pairs of dimples and springs provided on the straps for supporting the fuel rods, the dimple and the spring being disposed in facing relation to each other, on wall portions of the straps, which cooperate with each other to define the grid cells, the dimple and the spring projecting into the grid cell, said apparatus comprising: 13. An assembly apparatus according to claim 12, further comprising means for permitting the relative movement of said first and second support members and permitting the second support member to follow the movement of the first support member. |
abstract | A neutron reflector bolt fastening structure is disclosed in which even upon relaxation in the fastening forces thereof being generated in tie rods for divided stage portions as a result of neutron irradiation, it is possible to press the neutron reflector firmly against a core vessel. The neutron reflector bolt fastening structure includes: a neutron reflector which includes of a plurality of divided stage portions and situated in a core vessel in a reactor vessel; a plurality of tie rods for fixing the neutron reflector to the core vessel; and a plurality of bolts for exclusively fixing the lowermost stage portion of the plurality of stage portions of the neutron reflector to the core vessel. |
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claims | 1. A rotation apparatus usable with a control drum in a nuclear environment, the control drum having a shaft that is rotatable about an axis of rotation that is horizontal, a reflector portion situated on the shaft, an absorber portion situated on the shaft, and a motor which, when powered, is operable to move the shaft between an operational position wherein the reflector portion faces toward a core of the nuclear environment and a shutdown position wherein the absorber portion faces toward the core, the rotation apparatus comprising:a rotation mechanism which is structured to apply to the shaft in the operational position a force that is structured to rotate the shaft toward the shutdown position, the force being resisted by the motor to retain the shaft in the operational position when the motor is powered, the force not being resisted when the motor is unpowered; anda rotation management system that is structured to resist rotation of the shaft when the shaft is in the shutdown position. 2. The rotation apparatus of claim 1 wherein the rotation mechanism comprises a weight that is structured to be connected with the shaft and that is structured to be movable between a first position in the operational position of the shaft and a second position in the shutdown position of the shaft, the weight in the first position being vertically higher than in the second position and being movable by gravity from the first position to the second position when the motor is unpowered. 3. The rotation apparatus of claim 2 wherein the weight is a counterweight structured to be situated on the shaft such that a center of gravity of the counterweight is spaced from the axis of rotation, the counterweight situated on the shaft being structured to be movable with the shaft between the first position in the operational position of the shaft and the second position in the shutdown position of the shaft. 4. The rotation apparatus of claim 3 wherein the center of gravity is aligned in the vertical direction with the axis of rotation and disposed below the axis of rotation in the second position. 5. The rotation apparatus of claim 4 wherein the rotation management system comprises an electrically conductive structure situated on the shaft and an eddy current brake that comprises a number of magnets situated at opposite sides of the electrically conductive structure in the second position. 6. The rotation apparatus of claim 4 wherein the rotation mechanism further comprises a rotation initiator that comprises a pair of magnets, one of the magnets of the pair of magnets being structured to be situated on the control drum, the other of the magnets of the pair of magnets being structured to be situated on a support on which the control drum is disposed, the pair of magnets having their poles arranged to mutually oppose one another and being positioned to bias the control drum away from the operational position when the shaft is in the operational position, the bias being resisted by the motor when the motor is powered, the bias not being resisted when the motor is unpowered. 7. The rotation apparatus of claim 4 wherein the rotation mechanism further comprises a rotation initiator that comprises a spring that is structured to extend between the control drum and a support on which the control drum is disposed, the spring being positioned to bias the control drum away from the operational position when the shaft is in the operational position, the bias being resisted by the motor when the motor is powered, the bias not being resisted when the motor is unpowered. 8. The rotation apparatus of claim 3 wherein the rotation management system comprises at least one of a hard stop that is structured to be engaged by the control drum in the shutdown position and an eddy current brake that comprises a number of magnets. 9. The rotation apparatus of claim 1 wherein the rotation mechanism comprises a spring that is structured to apply the force to the shaft in the operational position to bias the shaft toward the shutdown position. 10. The rotation apparatus of claim 9 wherein the rotation management system comprises at least one of a hard stop that is structured to be engaged by the control drum in the shutdown position and an eddy current brake that comprises a number of magnets. 11. The rotation apparatus of claim 1 wherein the rotation management system comprises a lock that comprises a first portion and a second portion, the first portion being structured to be situated on one of the control drum and a support on which the control drum is disposed, the second portion being structured to be situated on the other of the control drum and the support, the lock being movable between a locked position wherein the first portion and the second portion are in a fixed relationship with one another and an unlocked position wherein one of the first portion and the second portion is movable with respect to the other of the first portion and the second portion. 12. The rotation apparatus of claim 11 wherein the shaft is in the shutdown position when the lock is in the locked position, the lock in the locked position being structured to resist movement of the shaft away from the shutdown position. 13. The rotation apparatus of claim 12 wherein the first portion is a bolt, wherein the second portion is a receptacle, and wherein the bolt is received in the receptacle in the locked position, the bolt being removed from the receptacle in the unlocked position. 14. The rotation apparatus of claim 13 wherein the receptacle is a cutout formed on the shaft, and wherein the bolt is structured to be situated on the support and to be movable on the support between a first location received in the receptacle and a second location removed from the receptacle. 15. The rotation apparatus of claim 14 wherein the lock further comprises an actuator that is structured to be disposed on the support, the bolt being disposed on the actuator, the actuator being structured to be operable to move the bolt between the first and second locations. 16. A control drum apparatus comprising the rotation apparatus of claim 1, the control drum apparatus being usable in a nuclear environment having a core and further comprising:a shaft that is rotatable about an axis of rotation that is horizontal;a control drum, the control drum having a reflector portion situated on the shaft and an absorber portion situated on the shaft; anda motor which, when powered, is operable to move the shaft between an operational position wherein the reflector portion faces toward the core and a shutdown position wherein the absorber portion faces toward the core. |
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052232115 | summary | BACKGROUND OF THE INVENTION This invention relates to a novel zirconium based alloy plate, a method of manufacturing this alloy plate, a fuel channel and a fuel assembly using this alloy plate. Zirconium is a material having high corrosion resistance and a small neutron absorption cross section and is therefore used for a reactor fuel assembly member. For this kind of use, Zr-Sn-Fe-Cr-Ni alloys called zircaloy-2 and zircaloy-4 are mainly used. If these alloys are used in a nuclear reactor for a long time, elongation, bending and deformation in particular directions occur as shown in FIG. 2, because (0001) planes thereof are oriented in the direction of plate thickness. If a bending deformation occurs in a fuel channel box, the space for driving the control rod is reduced. Such a phenomenon impedes the operation of the reactor. Also, if a bending deformation occurs, the distance to the fuel cladding tube is changed so that the ratio of water and uranium is locally changed, thereby causing a change in fission reactivity and, hence, abnormal heating. By this effect, corrosion of the fuel cladding tube is accelerated. Further, fuel damage may thereby be caused. To prevent bending deformation of the fuel channel box due to such non-uniformity of neutron exposure, uniformization of neutron exposure based on changing the fuel assembly loading position in the reactor core has been studied. This study, however, has not succeeded in preventing bending deformation. The reduction in the control rod driving space and the change in fission reactivity caused by bending deformation are major factors of limitation of the service life of the fuel channel box. The corrosion of the fuel channel box also is a factor of limitation of the service life. Japanese Patent Examined Publication Nos. 56-12310 and 60-44387 disclose a method for improving the corrosion resistance by a heat treatment based on quenching a Zr alloy from an .alpha.+.beta. phase temperature range or a .beta. phase temperature range. By this method, however, the bending deformation due to irradiation growth cannot be reduced for the reason described later, and this method provides no effective technique for limiting irradiation growth of zirconium alloy members. This conventional method uses a heat treatment for improving only the corrosion resistance of zirconium alloy members without changing the grain size, orientation and mechanical characteristics thereof. For this heat treatment, therefore, heating to an .alpha.+.beta. phase temperature range (800 to 980.degree. C.) rather than a .beta. phase temperature range (.gtoreq.980.degree. C.) in which coarsening of crystal grains occurs, followed by quenching, is effected. The crystal orientation, which is an important factor of limitation of irradiation growth, is not changed by this heat treatment and therefore no effective irradiation growth limiting technique is provided. Japanese Patent Unexamined Publication No.59-229475 discloses a method of orienting crystals so that an fl value is 0.15 to 0.5. This method also provides no irradiation growth limiting technique or rather promotes irradiation growth. SUMMARY OF THE INVENTION An object of the present invention is to provide a zirconium alloy member free from the above-mentioned elongation and bending deformation due to irradiation growth. Specifically, a main object of the present invention is to provide a long-life zirconium based alloy tubular member and a channel box in which the irradiation growth does not occur. According to the present invention, these objects can be achieved by making <0001> orientation of a hexagonal crystal of a zirconium alloy plate substantially perfectly random as shown in FIG. 1. According to one aspect of the present invention, there is provided a zirconium based alloy plate of low irradiation growth, containing not more than 5 wt % Sn and/or not more than 5 wt % Nb, and the balance Zr of not less than 90 wt %, the alloy plate having <0001> orientation of hexagonal crystal which orientation (Fr value) ranges from 0.20 to 0.5 with respect to direction perpendicular to the surface of the plate. According to another aspect of the present invention, there is provided a square tubular member made of a zirconium based alloy plate of low irradiation growth, containing not more than 5 wt % Sn and/or not more than 5 wt % Nb, and the balance Zr of not less than 90 wt %, the alloy member having <0001> orientation of hexagonal crystal in which <0001> orientation (Fr value) with respect to direction perpendicular to the surface of the tubular member ranges 0.20 to 0.50, another orientation (Ft value) with respect to longitudinal direction of the cylindrical member ranges 0.25 to 0.36, and still another orientation (Fl value) with respect to circumferential direction of the tubular member ranges 0.25 to 0.36. According to still another aspect of the present invention, there is provided a zirconium based alloy plate of low irradiation growth, containing not more than 5 wt % Sn and/or not more 5 wt % Nb, and the balance Zr of not less than 90 wt %, the alloy having .alpha. phase, and the grain size of the alloy being in the range of 50 to 500 .mu.m. According to a further aspect of the present invention, there is provided a zirconium based alloy member of low irradiation growth, containing not more than 5 wt % Sn and/or not more than 5 wt % Nb, and the balance Zr of not less than 90 wt %, the alloy having .alpha. phase, the alloy plate having <0001> orientation of hexagonal crystal in which <0001> orientation is substantially random, and strain occurring due to fast neutron irradiation of 2.times.10.sup.22 (n/cm.sup.3) being not more than 3.times.10.sup.-4. According to still a further aspect of the present invention, there is provided a method of producing a zirconium based alloy plate of low irradiation growth which contains not more than 5 wt % Sn and/or not more than 5 wt % Nb and the balance Zr of not less than 90 wt %, in which the alloy is heated in a .beta. single phase temperature range and is cooled, the method comprising the steps of retaining the alloy in the .beta. single phase temperature range in a short period of time so that <0001> orientation (Fr value) with respect to direction perpendicular to the surface of the alloy member is 0.20 to 0.40. The method of producing a zirconium based alloy plate of low irradiation growth in accordance with the present invention is characterized by comprising the step of retaining in a short period of time the alloy member in the .beta. single phase temperature range so that value of parameter P defined by P=(3.55+log t).times.log(T-980) where t is a retention period of time and T(h) is a retaining temperature (.degree. C.) is not less than 0.8, and quenching the alloy. According to still a further aspect of the present invention, there is provided a method of producing a square tubular member made of a low irradiation growth zirconium based alloy containing not more than 5 wt % Sn and/or not more than 5 wt % Nb and the balance Zr of not less than 90 wt %, comprising the steps of: locally induction-heating and keeping the tubular member in a .beta. single phase temperature range; and forcibly cooling the heated portion of the tubular member by cooling medium, the improvement comprising the steps of retaining the tubular member portion in a short period of time in the .beta. single phase temperature range so that <0001> orientation (Fr value) with respect to direction perpendicular to the surface of the square tubular member is 0.20 to 0.40; and then quenching the heated portion. According to a further aspect of the present invention, there is provided a method of producing a square tubular member made of a zirconium based alloy, comprising the steps of continuously induction-heating a portion of the tubular member locally while moving it relatively, and forcibly cooling the heated portion by a cooling medium, the improvement comprising the steps of inserting in the tubular member a mandrel made of a metal material having a thermal expansion coefficient larger than that of the alloy, and heating the tubular member from the outer surface of the tubular member while fixing at least both ends of the tubular member by the mandrels. According to a further aspect of the present invention, there is provided a fuel channel box formed of a square tubular member formed by welding two channel-shaped members made of zirconium based alloy, the alloy having <0001> orientation (Fr value ) with respect to direction perpendicular to the surface of the square tubular member of 0.20 to 0.50, the whole surface of the channel box being provided with an oxide layer formed by an autoclave treatment. According to a further aspect of the present invention, there is provided a fuel assembly comprising a fuel rod provided within a fuel cladding tube with fuel pellets, a channel box receiving a plurality of pieces of the fuel rods, a spacer for partitioning the fuel rods received in the channel box, and upper and lower tie plates disposed respectively at the upper and lower portions of the channel box, the channel box being made of a zirconium based alloy containing not more than 5 wt % Sn and/or not more than 5 wt % Nb and the balance Zr, the alloy having a texture that <0001> orientation (Fr value) with respect to direction perpendicular to the surface of the plate ranges from 0.20 to 0.50. According to a further aspect of the present invention, there is provided a method of using a nuclear fuel channel box made of a zirconium based alloy in which channel box a plurality of nuclear fuel rods are disposed, the alloy having a texture that <0001> orientation (Fr value) with respect to direction perpendicular to the surface of the tubular member ranges from 0.20 to 0.50, the nuclear fuel being exchanged during the use of the channel box at least two times. According to still a further aspect of the present invention, there is provided a method of using a fuel channel box formed of a tubular member formed by welding two channel-shaped members made of a zirconium based alloy, the alloy having a texture that <0001> orientation (Fr value) with respect to direction perpendicular to the surface of the tubular member ranges from 0.20 to 0.50, the whole surface of the channel box being provided with an oxide film formed by an autoclave treatment, the channel box being used so that the degree of burn-up on taking-out is not less than 32 Gwd/t or so that nuclear fuel is exchanged at least two times during the use thereof. According to still a further aspect of the present invention, there is provided a method of using a fuel assembly having a fuel assembly having a fuel rod having fuel pellets disposed within a fuel cladding tube, a channel box receiving a plurality of pieces of the fuel rods, a spacer for partitioning the fuel rods received in the channel box, and upper and lower lattice plates disposed respectively at the upper and lower portions of the channel box, the channel box being made of a zirconium based alloy containing not more than 5 wt % Sn and/or not more than 5 wt % Nb and the balance Zr of not less than 90 wt %, the alloy having a texture that <0001> orientation (Fr value) with respect to direction perpendicular to the surface of a plate ranges from 0.20 to 0.50, the channel box being used so that the degree of burn-up on taking-out is not less than 32 GWd/t or so that nuclear fuel is exchanged at least two times during the use thereof. According to a further aspect of the present invention, there is provided a method of using a nuclear fuel channel box made of a zirconium based alloy in which channel box a plurality of nuclear fuel rods are disposed, the alloy comprising hexagonal crystals having a texture that <0001> orientation (Fr value) with respect to direction perpendicular to the surface of the tubular member ranges from 0.20 to 0.50, the channel box being used so that the degree of burn-up on taking-out is not less than 32 GWd/t or so that it exposes neutron irradiation not less than 10.sup.22 n/cm.sup.2 during the use thereof. According to a further aspect of the present invention, there is provided a method of operating a nuclear reactor having within a reactor core a plurality of nuclear fuel channel boxes each formed of a tubular box made of a zirconium based alloy, comprising the steps of exchanging fuel after a predetermined period of operation time, and subsequently operating the reactor in a predetermined period of time, the alloy containing hexagonal crystals having a texture that <0001> orientation (Fr value) with respect to direction perpendicular to the surface of the tubular member ranges from 0.20 to 0.50, the channel boxes which are subjected to such operation as the degree of burn-up on taking-out is about 32 GWd/t or more being disposed in the same operation position as that of prior operation thereof, and fuel being exchanged at least two times during the use thereof. It is not only to exchange the channel box at least two times but also to use the channel box under a less deformation thereof for such period of time as a degree of burn-up on taking out becomes 32 GWd/t or more or for such a period of time as neutron irradiation at an amount of neutron exposure becomes 10.sup.22 n/cm or more. Specifically, the channel box can be effectively used at a high degree of burn-up of 38 GWd/t or higher, or 45 GWd/t or higher. |
abstract | A nuclear reactor core comprising fissile material is surrounded by a core former. The core former comprises one or more single-piece annular rings wherein each single-piece annular ring comprises neutron-reflecting material. In some embodiments the core former comprises a stack of two or more such single-piece annular rings. In some embodiments the stack of single-piece annular rings is self-supporting. In some embodiments the stack of single-piece annular rings does not include welds or fasteners securing adjacent single-piece annular rings together. A core basket may contain the nuclear reactor core and the core former, and in some embodiments an annular gap is defined between the core former and the core basket. In some embodiments the core former does not include welds and does not include fasteners. |
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052232107 | abstract | A liquid metal cooled nuclear fission reactor plant having a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during reactor shutdown, or heat produced during a mishap. This reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary system when rendered inoperable. |
abstract | An irradiating device and a method for controlling it are provided. The device comprises an electron accelerator and a scanning box connected to the electron accelerator, wherein the scanning box is provided with a target, an electron beam exit window positioned at left or right side of the target and a scanning magnet. The device integrates the functions of both the existing irradiating device outputting electron beams and those outputting X-rays. When the scanning magnet is in operation, the irradiating device outputs electron beams; and when the scanning magnet is not in operation, the irradiating device outputs X-rays. Therefore, the device is capable of outputting two radiation sources so as to meet requirements for radiation-processing articles with different sizes. |
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054901847 | claims | 1. A method for determining the power output of a pressurized water reactor having an excore detector system, said method comprising the steps of: determining at a base time thermal power output of said reactor; measuring detector current in said excore detector system including measuring detector current at said base time; measuring core inlet temperature including measuring core inlet temperature at said base time; periodically measuring core three-dimensional power distribution in said reactor including measuring core three-dimensional power distribution in said reactor at said base time; and determining present core power as the ratio of present detector current to detector current at said base time multiplied by said thermal power at said base time and corrected as a function of changes in said core inlet temperature and in said core three-dimensional power distribution since said base time. P.sub.r.sup.R =relative core power at said base time which is said thermal power determined at said base time; I=present detector current; I.sup.R =detector current at said base time; Q.sub.wa.sup.R =weighted average core three-dimensional power distribution at said base time; Q.sub.wa =most recent weighted average core three-dimensional power distribution; T.sub.i.sup.R =core inlet temperature at said base time; T.sub.i =present core inlet temperature; and K=a constant determined from the following relationship: ##EQU11## wherein K is calculated from measurements taken at two different core inlet temperatures. excore detector means adjacent said reactor for generating a detector current responsive to neutron flux produced by said reactor core; means for measuring core inlet temperature; means for determining core three-dimensional power distribution in said reactor core; and means determining present core power as the ratio of present detector current to detector current at a base time multiplied by thermal power calculated at said base time and corrected as a function of present core inlet temperature and core inlet temperature at said base time measured by said means for measuring core inlet temperatures and as a function of a most recent measurement of core three-dimensional power distribution in said reactor core and core three-dimensional power distribution at said base time measured by said means for measuring core three-dimensional power distribution. P.sub.r.sup.R =relative core power at said base time which is said thermal power determined at said base time; I=present detector current; I.sup.R =detector current at said base time; Q.sub.wa.sup.R =weighted average core three-dimensional power distribution at said base time; Q.sub.wa =most recent weighted average core three-dimensional power distribution; T.sub.i.sup.R =core inlet temperature at said base time; T.sub.i =present core inlet temperature; and K=a constant determined from the following relationship: ##EQU13## wherein K is calculated from measurements taken at two different core inlet temperatures. 2. The method of claim 1 wherein said present core power is corrected as a function of a difference between said core inlet temperature at said base time and present core inlet temperature, and as a function of a difference between said three-dimensional core power distribution at said base time and a most recent three-dimensional core power distribution. 3. The method of claim 2 wherein said step of correcting said present core power as a function of changes in core inlet temperature comprises calculating a constant from core power determined at two different core inlet temperatures and then subsequently correcting core power by an exponential term in which said constant is multiplied by a difference in core inlet temperature at said base time and present core inlet temperature. 4. The method of claim 3 wherein said present core power is determined from the relationship: ##EQU10## wherein: P.sub.r =present relative core power; 5. The method of claim 1 wherein said step of measuring said three-dimensional core power distribution comprises utilizing an incore power measurement system to measure said three-dimensional core power distribution. 6. The method of claim 1 wherein said step of measuring said three-dimensional core power distribution comprises measuring core exit temperatures across said reactor and using said core exit temperatures, core inlet temperatures, and excore detector currents to determine said three-dimensional core power distribution. 7. The method of claim 1 wherein said excore detector system includes at least one excore detector having a top detector measuring a top detector current and a bottom detector measuring a bottom detector current and wherein present core power is determined as an average between a top detector core power measurement made using said top detector current and a bottom detector core power measurement made using said bottom detector current. 8. The method of claim 7 wherein said detector system comprises a plurality of excore detectors each having a top detector and a bottom detector and wherein said present core power is determined as an average of said detector top core power measurement and said bottom core power measurement for all of said plurality of excore detectors. 9. A system for determining the power output of a pressurized water reactor system having a reactor with a reactor core with a core inlet through which reactor coolant enters for circulation through said reactor core, a steam generator through which reactor coolant circulated through said reactor core passes before it returns to said core inlet and feedwater means supplying feedwater to said steam generator and providing a feedwater flow measurement from which thermal power generated by said reactor is calculated; said monitoring system comprising: 10. The system of claim 9 wherein said excore detector means comprises plural excore detector means each generating a detector current and wherein said means determining present core power calculates present core power as an average of present core power determined from each of said detector currents. 11. The system of claim 10 wherein plural excore detector means comprises multiple detectors each having a top detector section generating a top detector current and a bottom detector section generating a bottom detector current and wherein said means determining present core power calculates present core power as an average of present core power determined from each of said top detector currents and bottom detector currents. 12. The system of claim 10 wherein said means determining present core power calculates said average of present core power from present core power determined for each detector means according to the relationship: ##EQU12## wherein: P.sub.r =present relative core power; 13. The method of claim 1 wherein said present pressurized water reactor core power is determined on-line. |
abstract | A multi-leaf collimator for a radiotherapy apparatus comprises at least one array of laterally-spaced elongate leaves, each leaf being driven by an associated motor connected to the leaf via a drive means so as to extend or retract the leaf in its longitudinal direction, the drive means comprising a sub-frame on which at least a subset of the motors are mounted, the sub-frame being mounted at a location spaced from the leaf array in a direction transverse to the lateral and longitudinal directions, and including a plurality of threaded drives disposed longitudinally, each being driven by a motor and being operatively connected to a leaf thereby to drive that leaf. |
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claims | 1. A process for estimating the concentration (C) of a chemical element in the primary coolant of a nuclear reactor, the reactor comprising means for injecting a solution for diluting the concentration of said chemical element below a predetermined limit into the primary coolant, means for injecting a concentrated solution of said chemical element having a predetermined concentration (C*) into the primary coolant, and a sensor able to measure a quantity (Cm) representing the concentration of said chemical element in the primary coolant, wherein the process is an iterative process comprising repeatedly in each time step k:a stage of acquisition of a quantity (qdk) representing the injected flow of the dilution solution in step k, a quantity (qk) representative of the injected flow of concentrated solution in step k, and a quantity (Cmk) representing the concentration of said chemical element measured by the sensor in the primary coolant;a stage of calculating an estimated value (Cek+1) for the concentration of said chemical element in the primary coolant in step k+1 based on the representative quantities (qdk, qck, Cmk) acquired in step k. 2. A process according to claim 1, wherein the calculation stage is carried out with the help of Kalman equations. 3. A process according to claim 2, wherein the stage of calculation in step k is carried out by considering in the Kalman equations a state parameter x=ln(C) when the quantity (qk) representing the injected flow of concentrated solution is below a predetermined limit and a state parameter x=ln(C*−C) when the quantity (qk) representing the injected flow of the concentrated solution is above said predetermined limit. 4. A process according to claim 3, wherein the calculation stage in step k is carried out by considering in the Kalman equations a measured parameter y=ln(Cm) when the quantity (cck)representing the injected flow of concentrated solution is below a predetermined limit and a state parameter y=ln(C*−Cm) when the quantity (qk) representing the injected flow is above said predetermined limit. 5. A process according to claim 4, wherein the equations used in the calculation stage are:xk+1/k=xk /k+uk uk=−(Δtk/Vol) qdk when the quantity (qck)representing the injected flow of concentrated solution in step k is below a predetermined limituk=−(Δtk/Vol) qck when the quantity (qck) representing an injected flow of concentrated solution in step k is above a predetermined limitPk+1/k=Pk/k+W xk+1/k+1=xk+1/kKk+1(yk+1−xk+1/k)Pk+1/k+1=(1−Kk+1)Pk+1/k Kk+1=Pk+1/k/(Pk+1/k+V)where xk/k is the value of state parameter x in step k determined from the data available in step k, Δtk is the duration of time step k, Vol is the volume of the primary circuit, Pk/k is the variance for the state parameter x in step k determined from the data available in step k, and W and V are predetermined constants. 6. A process according to claim 5, wherein the V/W ratio lies between 100 and 10000. 7. A process according to claim 1, wherein it comprises an initialisation stage in the course of which an initial estimated value (Ce0) for the concentration of said chemical element in the primary coolant is calculated directly from the quantity (Cm0) representing the concentration of said chemical element measured by the sensor in the primary coolant. 8. A process according to claim 1, wherein the chemical element is boron or a boron compound. 9. A process according to claim 1, wherein the nuclear reactor is a pressurised water nuclear reactor. |
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050733338 | abstract | A method of decontaminating radio nuclide-contaminated corrosion products, which are sparingly soluble in acid, from primary system surfaces in reactors of the pressurized water reactor type and the boiler reactor type with hydrogen dosage and similar, by oxidation and concurrent dissolution in an acidic decontamination solution of the acid soluble corrosion products obtained by said oxidation. The characteristic feature of the method is that said oxidation is performed with Ce.sup.4+ ions, ozone and chromic acid in the presence of perhalogen acid, preferably perchloric acid, at a pH below 3. |
claims | 1. A radiation-shielding assembly for a container having a radioactive material disposed therein, the assembly comprising:a body comprising a sidewall at least partially defining a cavity, the body defining an opening into the cavity, the body also comprising radiation shielding material;a cap adapted for releasable attachment to the body when the cap is in a first orientation relative to the body and for non-attached engagement with the body when the cap is in a second orientation relative to the body, radiation shielding material of the cap closing the opening in both the first orientation and in the second orientation;the cap including a magnetic portion operable to attract the base when the cap is in the first orientation, the cap being constructed to inhibit magnetic attraction of the cap to the base in the second orientation; andthe cap being operable to inhibit escape of radiation from the cavity of the assembly through the opening when the cap is adjacent the opening in the first orientation and in the second orientation. 2. The assembly of claim 1, wherein the opening is a first opening, the first opening being adjacent a first end of the body, the body defining a second opening adjacent a second end of the body, the first opening being of a first size, the second opening being of a second size greater than the first size, the assembly further comprising a base releasably attached to the body adjacent the second opening, the base comprising a base shielding element operable to limit escape of radiation from the assembly through the second opening when the base is attached to the body. 3. The assembly of claim 1, wherein the body comprises a top part and a bottom part removably interconnected with the top part, the bottom part having a closed end and an open end, the open end having an opening of a first size, the top part defining the opening of the body, the opening of the body being of a second size smaller than the first size, the top part being removable from the bottom part for loading and unloading a container into the cavity. 4. The assembly of claim 1, wherein the cap is adapted to be placed on a flat surface and to support the body above the surface when the cap is in the second orientation. 5. The assembly of claim 1, wherein the cap comprises at least one of a radiation absorbing material and a radiation reflecting material. 6. Use of the radiation-shielding assembly of claim 1 in eluting a radioisotope from a radioisotope generator. 7. The assembly of claim 1, wherein when the cap is in the second orientation, the attraction of the magnetic portion of the cap to the body is sufficiently attenuated so that a weight of the cap is sufficient to separate the cap from the body when one of the body and the cap is urged away from the other. 8. A method of using a radiation-shielding assembly, the method comprising:placing a container in a body of the radiation-shielding assembly, wherein the body comprises radiation shielding material;loading radioactive material into the container through a needle inserted into the container, the loading occurring while the container is in the body of the radiation-shielding assembly;releasably attaching a cap of the radiation-shielding assembly while the cap is in a first orientation relative to the body, wherein the releasably attaching is due, at least in part, to magnetic attraction between the cap and the body when the cap is in a first orientation, and wherein the releaseably attaching comprises closing an opening in the body of the radiation-shielding assembly with radiation shielding material of the cap;detaching the cap from the body after the releasably attaching, wherein the detaching comprises uncovering the opening;removing a radioactive material from within the body through the opening thereof while the opening is uncovered;non-attachedly engaging the body and the cap while the cap is in a second orientation relative to the body, wherein there is a substantial decrease in magnetic attraction between the cap and the body when the cap is in the second orientation than when the cap is in the first orientation, and wherein the non-attachedly engaging comprises closing the opening in the body of the radiation-shielding assembly with radiation shielding material of the cap to inhibit escape of radiation through the opening; anddisengaging the body and the cap to uncover the opening after the non-attachedly engaging. 9. The method of claim 8, wherein the loading comprises receiving a radioisotope from a radioisotope generator. 10. The method of claim 8, further comprising:transporting the body containing the container loaded with radioactive material from a first location to a second location while the cap is attached to the body in the first orientation. 11. The method of claim 10, wherein the first location is adjacent a radioisotope generator and the second location is adjacent a calibration system. |
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056298722 | abstract | A method and system for monitoring an industrial process and a sensor. The method and system include generating a first and second signal characteristic of an industrial process variable. One of the signals can be an artificial signal generated by an auto regressive moving average technique. After obtaining two signals associated with one physical variable, a difference function is obtained by determining the arithmetic difference between the two pairs of signals over time. A frequency domain transformation is made of the difference function to obtain Fourier modes describing a composite function. A residual function is obtained by subtracting the composite function from the difference function and the residual function (free of nonwhite noise) is analyzed by a statistical probability ratio test. |
abstract | A plurality of vessel holder means 14 are mounted around a revolving body 12 at an equal circumferential spacing. Each vessel holder means 14 includes two holders 26, 26 for carrying two vessels 4, 4 in vertical alignment. A path along which the revolving body 12 revolves contains an inversion interval C-D and an upright transfer interval D-B and A-C. Inversion means 16 which inverts the vessel holder means 14 about a tangential axis O1 is located within the inversion interval C-D while an electron beam irradiator 24 is located within the upright transfer interval A-C. Vessels which are fed at a vessel feed position A and carried by the holders 26 are subject to the irradiation of the electron beam at an electron beam irradiation position E, then inverted, and when it reaches the electron beam irradiation position E again, the opposite surface is subject to the irradiation of the electron beam. Vessels 4 are inverted again and then discharged at a vessel discharge position B. Using a single electron beam irradiator 24, the entire surface of the vessel 4 being conveyed can be sterilized by irradiation of the electron beam. |
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claims | 1. A method for predicting dose repeatability for an ion implantation, comprising:tuning an ion source to generate a tuned ion beam with desired beam current;obtaining beam current measurements from the tuned ion beam; andpredicting the dose repeatability for the ion implantation as a function of the beam current measurements using at least a standard deviation, an average, and a sample size of the beam current measurements. 2. The method according to claim 1, wherein the obtaining beam current measurements includes taking measurements along the beamline. 3. The method according to claim 1, further comprising determining whether the predicted dose repeatability is less than a target dose repeatability for the ion implantation. 4. The method according to claim 3, further comprising tuning the ion source to generate an ion beam with desired beam noise and/or beam current if the predicted dose repeatability is greater than the target dose repeatability for the ion implantation. 5. The method according to claim 3, further comprising directing the ion beam into an end station that is configured to receive a substrate during the ion implantation if the predicted dose repeatability is less than the target dose repeatability for the ion implantation. 6. The method according to claim 5, further comprising obtaining beam current measurements from the end station. 7. The method according to claim 6, further comprising predicting the dose repeatability for the ion implantation as a function of the beam current measurements obtained from the end station. 8. The method according to claim 7, further comprising determining whether the predicted dose repeatability is less than the target dose repeatability for the ion implantation. 9. The method according to claim 8, further comprising tuning the ion source to generate an ion beam with desired beam noise and/or beam current if the predicted dose repeatability is greater than the target dose repeatability for the ion implantation. 10. The method according to claim 8, further comprising performing the ion implantation of the substrate within the end station if the predicted dose repeatability is less than the target dose repeatability for the ion implantation. 11. A method for controlling an ion implantation of a substrate according to predicted dose repeatability, comprising:tuning an ion source to generate a tuned ion beam suitable for performing the ion implantation of the substrate;obtaining beam current measurements from the tuned ion beam;determining the predicted dose repeatability for the ion implantation as a function of a standard deviation, an average, and a sample size of the beam current measurements; andcontrolling the ion implantation of the substrate as a function of the predicted dose repeatability. 12. The method according to claim 11, wherein the obtaining beam current measurements includes taking measurements along the beamline and an end station configured to receive the substrate during the ion implantation. 13. The method according to claim 11, wherein the tuning of the ion source comprises adjusting beamline element settings to generate an ion beam with desired beam properties. 14. The method according to claim 13, wherein the desired beam properties include noise and beam current. 15. The method according to claim 11, wherein the controlling of the ion implantation comprises retuning the ion source to attain a predicted dose repeatability that conforms to a target dose repeatability for the ion implantation. 16. A system for predicting dose repeatability for an ion implantation of a substrate, comprising:a tuner configured to tune an ion source to generate an ion beam suitable for performing the ion implantation of the substrate;a beamline monitor configured to obtain beam current measurements from the tuned ion beam; anda controller configured to predict the dose repeatability for the ion implantation as a function of a standard deviation, an average, and a sample size of the beam current measurements obtained by the beamline monitor. 17. The system according to claim 16, wherein the controller is configured to determine whether the predicted dose repeatability is in conformance with a target dose repeatability for the ion implantation. 18. The system according to claim 17, wherein the controller is configured to direct the ion beam from the ion source into an end station configured to receive the substrate during the ion implantation if the predicted dose repeatability is in conformance with the target dose repeatability. 19. The system according to claim 18, further comprising an end station monitor configured to obtain beam current measurements from the end station. 20. The system according to claim 19, wherein the controller is configured to predict the dose repeatability for the ion implantation as a function of the beam current measurements obtained from the end station by the end station monitor. 21. The system according to claim 20, wherein the tuner is configured to retune the ion source to generate an ion beam with desired beam noise and/or beam current if the predicted dose repeatability deviates from a target dose repeatability for the ion implantation. 22. An ion implanter, comprising:an ion source configured to generate an ion beam;a magnet configured to bend the path of the ion beam;an end station configured to receive the ion beam from the magnet for ion implantation of a substrate within the end station; anda controller configured to control the ion implantation of the substrate as a function of predicted dose repeatability, wherein the controller predicts dose repeatability from at least a standard deviation, an average, and a sample size of beam current measurements obtained from the ion beam after tuning of the ion source. 23. A computer-readable medium storing computer instructions, which when executed by a computer system enables an ion implanter to control an ion implantation of a substrate according to predicted dose repeatability, the computer instructions comprising:tuning an ion source to generate an ion beam suitable for performing the ion implantation of the substrate;obtaining beam current measurements from the ion beam;predicting the dose repeatability for the ion implantation as a function of a standard deviation, an average, and a sample size of the beam current measurements; andcontrolling the ion implantation of the substrate as a function of the predicted dose repeatability. 24. The computer-readable medium according to claim 23, wherein the obtaining beam current measurements includes instructions for taking measurements along the beamline and an end station configured to receive the substrate for the ion implantation. 25. The computer-readable medium according to claim 23, wherein the controlling of the ion implantation comprises instructions for retuning the ion source to attain a predicted dose repeatability that conforms to a target dose repeatability for the ion implantation. |
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abstract | A nuclear power plant is provided including a BWR, a reactor cooling system cooling the BWR, an HWC hydrogen injection system connected to the reactor cooling system and an alcohol injection system connected to the reactor cooling system. Methods for providing methanol and hydrogen are also provided. |
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description | This application is a divisional of patent application Ser. No. 12/028,215, filed Feb. 8, 2008, and claims the benefit of the filing date thereof 1. Field of the Invention The present invention relates generally to nuclear reactors and, more particularly, to an improved neutron absorber material contained in gray rod assemblies of gray rod control assemblies (GRCAs). 2. Description of the Prior Art The fuel assemblies of modern reactor cores typically employ two types of rod control assemblies to control reactivity, rod cluster control assemblies (RCCAs) and gray rod control assemblies (GRCAs). Both consist of a plurality of neutron-absorbing rods fastened at their top ends to a common hub or spider assembly. The body of the rods generally comprises a stainless steel tube which encapsulates a neutron-absorbing material, such as a pure silver absorber material or a silver-indium-cadmium alloy absorber material, and the rods are slid within tubular guide thimble tubes of the fuel assembly with a control drive mechanism near the top of the spider assembly operating to control the movement of the rods within the thimble tubes. In this manner, the controlled insertion and extraction of the rods generally controls the rate of reactor power produced. The power produced by the reactor of a nuclear power plant is generally controlled by raising or lowering control rod assemblies within the reactor core, and the change in reactor power output required in order to accommodate a change in the demand for electrical output from the electrical power plant is commonly referred to as load follow. As described, for example, in U.S. Pat. No. 4,079,236, load follow presents many operating issues. For instance, in a pressurized water reactor (PWR) during load follow, reactivity must be controlled and axial power distribution changes in the core in response to the power level change, must be addressed. Typically, GRCAs are used in load follow maneuvering because they comprise reduced worth control rods, commonly referred to in the art as “gray” rods. Gray rods are known to provide a mechanical shim (MSHIM) reactivity mechanism as opposed to a chemical shim, which requires changing the concentration of soluble boron in the reactor coolant. Thus, the use of gray rods minimizes the need for processing the primary reactor coolant on a daily basis and, therefore, greatly simplifies operations. More specifically, GRCA designs typically consist of twenty-four rodlets fastened at their top ends to the spider. Of the twenty-four rodlets within the cluster, only four rods are absorber rods, and the neutron-absorber material encapsulated within the absorber rods typically consists of about 80% silver, about 15% indium, and about 5% cadmium. Such a design poses several disadvantages. Among the disadvantages of known GRCA designs, is the fact that indium and cadmium have relatively large neutron cross-sections, which result in their depletion over a relatively short period of time. Silver depletes somewhat more slowly than indium and cadmium, and ultimately transmutes into other non-absorbing isotopes of cadmium. As a result of continued decrease in the rod worth, the GRCAs become less effective in controlling the reactivity of the reactor during load follow. In addition, elemental transmutation of silver and indium to other metals leads to changes in absorber material properties and excessive absorber swelling, which has been a recognized problem in the industry for many years. This undesirably leads to frequent GRCA replacement. A second disadvantage relates to changes in the local rod power for fuel rods which are adjacent to the four guide thimbles that contain the absorber rods. Specifically, because the absorber material is localized to four rodlets, a rapid change in power, commonly referred to as the delta-power of the fuel rods, occurs, for example, during a rod pull. A rod pull is the process of extracting the GRCA from the fuel assembly, and in GRCA designs it results in a delta-power spike. There exists a need, therefore, for an improved neutron absorber material for gray rod assemblies which overcomes the aforementioned disadvantages typically found in known GRCAs. This need and others are satisfied by the present invention, which is directed to an improved neutron absorbing material for gray rod control assemblies (GRCAs), which is mechanically self-supporting up to substantially higher temperatures than those at which pure silver or silver-indium-cadmium alloy retains its shape, while being spatially uniform in its ability to absorb neutrons. In one aspect of the present invention, there is provided a gray rod assembly for a gray rod control assembly of a nuclear reactor. The nuclear reactor includes a number of fuel assemblies each having a plurality of elongated nuclear fuel rods supported in an organized array by a number of substantially transverse support grids, and a plurality of guide thimbles disposed through the support grids and along the fuel rods. The gray rod control assembly includes a spider assembly having a plurality of radially extending flukes and being structured to move each gray rod assembly within one of the guide thimbles in order to control the rate of power produced by the nuclear reactor. The gray rod assembly comprises an elongated tubular member having a first end, a second end, an inner diameter, and a length; a first end plug coupled to the first end of the elongated tubular member, and being structured to facilitate insertion of the elongated tubular member into one of the guide thimbles of the fuel assembly; a second end plug coupled to the second end of the elongated tubular member, and being structured to be coupled to one of the radially extending flukes of the spider assembly of the gray rod control assembly; and a neutron absorber comprised of a matrix of refractory metal fabricated to be porous into which a neutron absorbing metal or metal alloy is infused. The neutron absorber is disposed as a plurality of segments within most of the elongated tubular member, having a diameter which is relatively equivalent in diameter to the elongated tubular member, and a length which is shorter than the length of the elongated tubular member, in order to minimize the exposed surface area of the neutron absorber to radiation when the tubular member is inserted into the thimble and to allow the tubular member to flex, if necessary. The neutron absorber is distributed among a plurality of the gray rod assemblies. In another aspect of the present invention, there is provided a gray rod control assembly for a nuclear reactor. The nuclear reactor includes a plurality of fuel assemblies each having a plurality of elongated nuclear fuel rods supported in an organized array by a number of substantially transverse support grids, and a plurality of guide thimbles disposed through the support grids and along the fuel rods. The gray rod control assembly comprises a spider assembly having a plurality of radially extending flukes; and a plurality of gray rod assemblies coupled to the flukes of the spider assembly, the spider assembly being structured to move each gray rod assembly within one of the guide thimbles in order to control the rate of power produced by the nuclear reactor. Each of the gray rod assemblies comprises an elongated tubular member having a first end, a second end, an inner diameter, and a length; a first end plug coupled to the first end of the elongated tubular member, and being structured to facilitate insertion of the elongated tubular member into one of the guide thimbles of the fuel assembly; a second end plug coupled to the second end of the elongated tubular member, and being structured to be coupled to one of the radially extending flukes of the spider assembly; and a neutron absorber. The neutron absorber may comprise a matrix of refractory metal fabricated to be porous into which a neutron absorbing metal or metal alloy is infused. The neutron absorber is disposed as a plurality of segments within the elongated tubular member generally toward the first end. The neutron absorber has a diameter that is relatively equivalent in diameter to the elongated tubular member, and a length that is substantially shorter than the length of the elongated tubular member. The neutron absorber is distributed among a plurality of the gray rod assemblies. In a further aspect of the present invention, there is provided a nuclear reactor, comprising a plurality of elongated nuclear fuel rods each having an extended axial length; a number of substantially transverse support grids spaced along the axial length of the fuel rods in order to hold the fuel rods in an organized array; a plurality of guide thimbles disposed through the support grids and along the fuel rods; and a gray rod control assembly including a spider assembly having a plurality of radially extending flukes, and a plurality of gray rods assemblies coupled to the flukes, the advanced gray rod control assembly being structured to move each of the gray rod assemblies within one of the guide thimbles in order to control the rate of power produced by the nuclear reactor. Each of the gray rod assemblies comprises an elongated tubular member having a first end, a second end, an inner diameter, and a length, a first end plug coupled to the first end of the elongated tubular member, the first end plug being tapered in order to facilitate insertion of the elongated tubular member into one of the guide thimbles of the fuel assembly, a second end plug coupled at one end to the second end of the elongated tubular member, and at the other end to one of the radially extending flukes of the spider assembly, and a neutron absorber. The neutron absorber may comprise a matrix of refractory metal fabricated to be porous into which a neutron absorbing metal or metal alloy is infused. The neutron absorber is disposed as a plurality of segments generally filling most of the elongated tubular member. The neutron absorber has a diameter that is relatively equivalent to the inner diameter of the elongated tubular member. The neutron absorber is distributed among a plurality of the gray rod assemblies. The neutron absorber is comprised preferably of between about 40% to about 80% refractory metal and between about 20% to about 60% metal or metal alloy, more preferably of between about 50% to about 70% refractory metal and between about 30% to about 50% metal or metal alloy, and most preferably of about 65% refractory metal and about 35% metal or metal alloy. Suitable refractory metals of the present invention include, without limitation, molybdenum, tungsten, niobium or zirconium. An exemplary neutron absorber of the present invention includes, for example, a matrix of porous molybdenum as a refractory metal that is infused with silver or a silver-indium-cadmium alloy in the pore network of the refractory metal. The porous matrix of the refractory metal is accomplished, for example, by sintering. The neutron absorber material of the present invention may be shaped, for example and without limitation, as cylindrical pellets such as right circular cylindrical pellets. For simplicity of disclosure, the invention will be described with reference to the pressurized water reactor (PWR) core design which is commercially known under the designation AP1000. The AP1000 is a Westinghouse Electric Company LLC reactor design. Westinghouse Electric Company LLC has a place of business in Monroeville, Pa. Reference to the AP1000 is provided for illustrative example purposes only and is not limiting upon the scope of the invention. It will, therefore, be appreciated that the exemplary GRCA design of the invention has application in a wide variety of other reactor designs. Directional phrases used herein, such as, for example, upper, lower, top, bottom, left, right, and derivatives thereof, relate to the orientation of the elements shown in the drawings and are not limiting upon the claims unless expressly recited therein. As employed herein, the statement that two or more parts are “coupled” together shall mean that the parts are joined together either directly or joined through one or more intermediate parts. As employed herein, the term “number” shall refer to one and more than one (i.e., a plurality). Fuel Assembly Referring now to the drawings and particularly to FIG. 1, there is shown an elevational view of a nuclear reactor fuel assembly, represented in vertically shortened form and being generally designated by reference numeral 10. The fuel assembly 10 is the type used in a pressurized water reactor (PWR) and has a structural skeleton which, at its lower end, includes a bottom nozzle 12 for supporting the fuel assembly 10 on a lower core support plate 14 in the core region of the nuclear reactor (not shown), a top nozzle 16 at its upper end, and a number of guide tubes or thimbles 18 which extend longitudinally between and are rigidly attached at opposite ends, to the bottom and top nozzles 12 and 16. The fuel assembly 10 further includes a plurality of transverse grids 20 axially-spaced along and mounted to the guide thimble tubes 18 and, an organized array of elongated fuel rods 22 transversely-spaced and supported by the grids 20. The assembly 10 also has an instrumentation tube 24 located in the center thereof and extending between and mounted to the bottom and top nozzles 12 and 16. In view of the foregoing arrangement of parts, it will be understood that the fuel assembly 10 forms an integral unit capable of being conveniently handled without damaging the assembly parts. As previously discussed, the array of fuel rods 22 in the fuel assembly 10 are held in spaced relationship with one another by the grids 20 which are spaced along the fuel assembly length. Each fuel rod 22 includes nuclear fuel pellets 26 and is closed at its opposite ends by upper and lower end plugs 28 and 30. The pellets 26 are maintained in a stack by a plenum spring 32 disposed between the upper end plug 28 and the top of the pellet stack. The fuel pellets 26, composed of fissile material, are responsible for creating the reactive power of the reactor. A liquid moderator/coolant such as water or water containing boron, is pumped upwardly through a plurality of flow openings in the lower core plate 14 to the fuel assembly. The bottom nozzle 12 of the fuel assembly 10 passes the coolant upwardly through the guide thimbles 18 and along the fuel rods 22 of the assembly in order to extract heat generated therein for the production of useful work. To control the fission process, a number of control rods 33 without neutron absorber and control rods 34 with neutron absorber are reciprocally moveable in the guide thimbles 18 located at predetermined positions in the fuel assembly 10. A spider assembly 39 positioned above the top nozzle 16 supports the control rods 33, 34. FIGS. 2A and 2B show the control rod assembly 36 after it has been removed from the fuel assembly 10 of FIG. 1. Generally, the control assembly 36 has an internally threaded cylindrical member 37 with a plurality of radially-extending flukes or arms 38, which comprise the spider assembly 39, best shown in FIG. 2B. As previously discussed, each arm 38 is interconnected to the control rods 33 without neutron absorber and control rods 34 with neutron absorber, such that the control rod assembly 36 is operable to move the control rods 33, 34 vertically within the guide thimbles 18 (FIG. 1) to thereby control the fission process in the fuel assembly 10 (FIG. 1), all in a well known manner. With the exception of the exemplary control rod assembly which comprises a gray control rod assembly (GRCA) 36 having gray rod assemblies 34 with improved neutron absorbing material, which will now be discussed, all of the foregoing is old and generally well known in the art. Improved GRCA Continuing to refer to FIGS. 2A and 2B, the general control rod configuration will now be described. As previously stated, in order to take advantage of the MSHIM capabilities afforded by low worth or gray rods, known control rod assemblies, such as the existing control assemblies for the Westinghouse Electric Company LLC AP1000 reactor, employ GRCAs. However, while the GRCA design for the current AP1000 reactor design has twenty-four rods which are generally configured as shown in FIG. 2B, as mentioned hereinbefore, twenty of the twenty-four rods are stainless steel (e.g., without limitation, SS-304) water displacing rods and only four of the rods are neutron-absorber rods. Therefore, essentially all of the neutron absorber material is localized and isolated in only four rod locations within the GRCA. Additionally, in the existing AP1000 design, the absorber material comprises an Ag—In—Cd absorber consisting of about 80% silver, about 15% indium, and about 5% cadmium. This absorber material is consistent with known standard full-strength rod cluster control assemblies (RCCAs), in which all twenty-four rods are Ag—In—Cd. As noted, indium and cadmium are known to rapidly deplete. RCCAs spend a minimal amount of time in the core during power operation. Therefore, such depletion is not an issue. However, for the AP1000 mechanical shim (MSHIM) operation, for example, the GRCAs are expected to reside in the core for up to half of the operating cycle. Under these operating conditions, the existing GRCA design would need to be replaced periodically due to rapid absorber depletion. As will be discussed in detail herein, among other benefits, the improved GRCA design of the invention overcomes this rapid depletion disadvantage and also substantially avoids the undesirable local power spike experienced when traditional GRCA having four gray rod assemblies with neutron absorbing material is pulled from the core. Two-dimensional multi-group transport theory simulations also demonstrate that the exemplary neutron absorber of the invention, comprised of, for example and without limitation, a molybdenum refractory metal infused with silver, compared to gray rod designs composed of thin wires of pure silver metal or Ag—In—Cd surrounded by a steel spacer sleeve and a clad with approximately equal reactivity worth, is superior both in terms of depletion lifetime and intra-assembly peaking. Depletion calculations indicate that approximately one-third of the silver in the exemplary silver-molybdenum neutron absorber of the invention transmute to cadmium-108 or -110 at the end of its targeted lifetime. The molybdenum percentage, however, remains essentially constant over the targeted life of the gray rod assembly 34 (FIGS. 2A, 2B), with no significant quantities of other chemical species produced due to irradiation of the molybdenum. Hence, the material composition of the irradiated silver-molybdenum neutron absorber of the invention is expected to remain fairly similar to the initial composition. In contrast, depletion calculations for small diameter applications of Ag—In—Cd alloy indicate that the relative changes in material composition due to irradiation are significantly larger in these materials for the same targeted lifetime. A further understanding of the aforementioned absorber depletion issue will be had with reference to FIGS. 3 and 4, which show the gray rod assembly 34 with neutron absorber 110 of the invention. As shown in FIG. 3, the gray rod assembly 34 generally includes a first end 40 which, as oriented in the core (FIG. 1), is the bottom end, and a second end 42 (e.g., top end from the perspective of FIG. 1). The first or bottom end 40 has a tapered end plug 44. Such tapered design facilitates guided insertion of the rod 34 into the thimble tube 18 (FIG. 1) of the fuel assembly 10 (FIG. 1). The second or top end 42 has a top end plug 46 which is structured to engage and secure to the spider assembly 39 (best shown in FIG. 2A) in a known manner (e.g., without limitation, a complementary male/female fastening arrangement). An elongated tubular member 48 extends between the top and bottom end plugs 46, 44. The exemplary tubular member 48 is a stainless steel tube made from 304-stainless steel, although tubes made from other known or suitable alternative materials are contemplated. In the example shown and discussed herein, the inner diameter 50 of the tube 48 (FIG. 4) is about 0.38 inches (0.97 centimeters). However, it will be appreciated that the concepts of the invention are equally applicable for gray rod assemblies 34 having any suitable inner diameter for use in a wide variety of reactors. The neutron absorber material 110 comprises a matrix of porous refractory metal infused with a neutron absorbing metal or metal alloy. The refractory metal is fabricated, for example, by compacting and sintering metal powder so as to result in a continuous pore network, which then is infused with a neutron absorbing metal or metal alloy. The neutron absorber 110 preferably is between about 40% to about 80% refractory metal and between about 20% to about 60% neutron absorbing metal or metal alloy, more preferably between about 50% to about 70% refractory metal and between about 30% to about 50% neutron absorbing metal or metal alloy, and most preferably about 65% refractory metal and about 35% neutron absorbing metal or metal alloy. The refractory metal may be, for example, molybdenum, tungsten, niobium or zirconium. As discussed above, preferably, the refractory metal is molybdenum and the metal that is infused in the refractory metal preferably is, for example, silver. Alternatively, a neutron absorbing metal alloy such as, for example, Ag—In—Cd, may be infused in the refractory metal. The neutron absorber material 110 generally is disposed within most of the tube 48 in a plurality of segments 58 therein, shown not in scale in FIG. 3, in which each segment 58 comprises the neutron absorber 110, preferably in the form, for example, of cylindrical pellets. Segmentation of the neutron absorber 110 in the elongated tubular member 48 allows for the elongated tubular member 48 to be flexible so as to reduce frictional forces between the elongated tubular member 48 and the thimble tubes 18, which reduces the likelihood of incomplete insertion of the gray rod assemblies 34 in the thimble tubes 18. As best shown in the cross-sectional view of FIG. 4, the diameter 54 of the exemplary neutron absorber material 110 is relatively equivalent in diameter to the inner diameter 50 of the rod tube 48. The length 56 of the absorber 110 in the example of FIG. 3 is about 166 inches (421.64 centimeters). Although, as with the other dimensions of the gray rod 34, this measurement could vary without departing from the scope of the invention. The refractory metal of the invention serves as a structural component for the neutron absorbing material infused therein, i.e., the silver metal or silver-indium-cadmium metal alloy is infused in the refractory metal in relatively small amounts in the pore network of the refractory metal. The exemplary gray rod assembly 34 of the invention provides an extended nuclear lifetime through use of the exemplary neutron absorber 110. This is due to the low percentage of absorber metal, for example, about 35% silver, infused in the porous matrix of refractory metal, for example, about 65% molybdenum. Specifically, the low percentage of absorber metal resists bulk boiling during conditions of high local power density, absorber swelling and resultant clad cracking. The overall GRCA 36 design of the invention also generally improves fuel rod 22 linear heat rate change margins during GRCA 36 maneuvers. In an embodiment, of the twenty-four rods 33, 34 in the exemplary GRCA 36, about, for example, twelve rods 34 contain the exemplary neutron absorber 110 and the remaining rods 33 do not contain the exemplary neutron absorber 110, as opposed to localizing the absorber in only four rods, as in the existing AP1000 design discussed hereinbefore. However, it will be appreciated that the concepts of the invention are equally applicable for gray rods 33, 34 having any other suitable arrangement of neutron absorber-containing rods 34 and rods 33 without neutron absorber. In addition, the neutron absorber 110 may be distributed evenly over about the rods 34, which reduces the change in local fuel rod power (delta-power) when the GRCA 36 is removed from the core, which in turn improves operating margins. Further, distributing the absorber material 110 evenly over the rods 34 reduces the amount of absorber 110 in each rod 34, which reduces the amount of heat generated in each rod 34 and resists the risk of bulk boiling in the thimbles 18 under high local power density conditions. The exact reduction in amount of absorber material 110, as compared with the four Ag—In—Cd absorbers of the current design, is not meant to be limiting upon the invention. In view of the foregoing, the exemplary gray rod control assembly 36 has been redesigned to address and substantially overcome the aforementioned disadvantages in the art by including an entirely different absorber material 110 comprising a porous matrix of refractory metal infused with a neutron absorbing metal or metal alloy, a reduced amount of absorber 110 in the rods 34 containing the neutron absorber, distribution of the neutron absorber 110 in discrete segments 58 of the rods 34, and distribution of the neutron absorber 110 evenly among, for example, about twelve rods 34 of the twenty-four rods 33, 34. |
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claims | 1. A pre-patient collimator for controlling x-ray exposure during a scan with a computed tomography system, the computed tomography system including a patient table translatable along a z-axis, an x-ray source and a detector array, said collimator comprising:a first cam configured to shutter an x-ray fan beam, generated by the x-ray source, in a first direction;a second cam located on an opposite side of a focal point of the x-ray source from the first cam, said second cam configured to shutter the x-ray fan beam in a second direction, the second direction being opposite from the first direction; anda cam drive configured to position at least one of said first cam and said second cam during the scan at a rate proportional to a translation speed of the table. 2. A collimator in accordance with claim 1 wherein said first cam and said second cam are fabricated from a x-ray absorbing material, and said first cam and said second cam are configured to block a portion of an x-ray beam directed towards an object to be imaged to adjust the thickness of the fan beam. 3. A collimator in accordance with claim 2 wherein said cam drive is configured to independently position each said first cam and said second cam. 4. A collimator in accordance with claim 2 wherein said cam drive comprises at least one cam motor. 5. A collimator in accordance with claim 4 wherein said cam drive comprises a cam motor for each said first cam and said second cam. 6. A collimator in accordance with claim 2 wherein said cam drive is configured to modify a fan beam thickness. 7. A system for reducing x-ray exposure in a computed tomography system, the computed tomography system including a patient table translatable along a z-axis, an x-ray source, a detector array comprising a plurality of detector cells extending in the z-axis, and a pre-patient collimator comprising a shutter cam and a tracking cam, said system configured to:determine a speed of a patient table along the z-axis; andposition at least one of the shutter cam and the tracking cam during a scan at a rate proportional to the determined table speed to define a thickness of a fan beam to be directed towards an object to be imaged. 8. A system in accordance with claim 7 wherein said cams are configured to block a portion of an x-ray beam emitted by the x-ray source to adjust the thickness of the fan beam. 9. A method controlling x-ray dosage in a computed tomography system including a patient table translatable along a z-axis and a collimator having a first cam and a second cam that define an aperture, said method comprising:opening the first cam at a rate proportional to a translation speed of the table; andclosing the second cam at a rate proportional to the translation speed of the table. 10. A method in accordance with claim 9 further comprising determining a speed of translation of the table along the z-axis. 11. A method in accordance with claim 9 wherein opening the first cam comprises initiating opening the first cam when a leading edge of an object of interest intersects an edge of an x-ray beam defined by the first cam. 12. A method in accordance with claim 9 wherein opening the first cam comprises opening the first cam at a velocity determined using: velocity = ( source_to _cam source_to _iso + SFOV 2 ) * table_speed . 13. A method in accordance with claim 9 wherein opening the first cam comprises initiating opening the first cam when a leading edge of an object of interest is coincident with an edge of an x-ray beam defined by the first cam at a velocity determined by: velocity = ( source_to _cam source_to _iso - SFOV 2 ) * table_speed . 14. A method in accordance with claim 9 wherein closing the second cam comprises initiating closing the second cam when a trailing edge of an object of interest intersects an edge of an x-ray beam defined by the second cam. 15. A method in accordance with claim 9 further comprising fully closing the first cam prior to the beginning of a scan. 16. A method in accordance with claim 9 further comprising fully opening the second cam prior to the beginning of a scan. 17. A method in accordance with claim 9 further comprising maintaining the first cam at a position Bcam for a time t0<t<tB, wherein to represents a time when the first view is collected, tB represents a time for a leading edge of an object of interest to intersect the beam edge defined by the first cam when the first cam is fully closed, and wherein Bcam is a fully closed position of the first cam. 18. A method in accordance with claim 9 further comprising maintaining the first cam in a position defined by, Bcam+(t−tB)*vBC, for a time, tB<t<tC, wherein tB represents a time for a leading edge of an object of interest to intersect the beam edge defined by the first cam when the first cam is fully closed, Bcam is a fully closed position of the first cam, and vBC is a velocity of the first cam. 19. A method in accordance with claim 9 further comprising maintaining the first cam in a position defined by, Bcam+(t−tB)*vBC, for a time, tB<t<tC, wherein tB represents a time when a leading edge of an object of interest intersects the beam edge defined by the first cam when the first cam is fully closed, tC represents a time when the object of interest crosses the centerline of the detector (C) Bcam is a fully closed position of the first cam, and vBC is a velocity of the first cam. 20. A method in accordance with claim 9 further comprising maintaining the first cam in a position defined by, Bcam+(tC−tB)*vBC+(t−tC)*vCT, for a time, tC<t<tT, wherein tB represents a time when a leading edge of an object of interest intersects the beam edge defined by the first cam when the first cam is fully closed, tC represents a time when the object of interest crosses the centerline of the detector (C), tT represents a time when the object of interest intersects the beam edge defined by the first cam when the first cam is fully open, Bcam is a fully closed position of the first cam, and vBC and vCT are velocities of the first cam. 21. A method in accordance with claim 9 further comprising maintaining the first cam in a fully open position, Bcam, for a time, t>tT, wherein tT represents a time when the leading edge of the object of interest intersects the beam edge defined by the first cam when the first cam is fully open. 22. A method in accordance with claim 9 further comprising maintaining the second cam in a fully open position, Tcam, for a time, t<tT′, wherein tT′ represents a time when a trailing edge of the object of interest intersects the beam edge defined by the second cam when the second cam is fully open. 23. A method in accordance with claim 9 further comprising maintaining the second cam in a position defined by, Tcam+(tC′−tT′)*vTC′, for a time, tT′>t>tC′, wherein tC′ represents a time when the trailing edge of the object of interest crosses the centerline of the detector (C), tT′ represents a time when the trailing edge of the object of interest intersects the beam edge defined by the second cam when the second cam is fully open, Tcam is a fully open position of the second cam, and vTC is a velocity of the second cam. 24. A method in accordance with claim 9 further comprising maintaining the second cam in a position defined by, Tcam+(tB′−tC′)*vB′C′+(tC′−tT′)*vTC′, for a time, tC′<t<tB′, wherein tB′ represents a time when a leading edge of an object of interest intersects the beam edge defined by the first cam when the first cam is fully closed, tC represents a time when the object of interest crosses the centerline of the detector (C) Bcam is a fully closed position of the first cam, and vBC is a velocity of the first cam. 25. A method in accordance with claim 9 further comprising maintaining the first cam at a position Tcam for a time t>tB′, wherein tB′ represents a time for the trailing edge of the object of interest to intersect the beam edge defined by the second cam when the second cam is fully closed, and wherein Tcam is a fully closed position of the second cam. |
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claims | 1. A method of inspecting welding of a sealed closure plug of a fuel element for a nuclear reactor, the fuel element having a tubular sleeve enclosing a plurality of nuclear fuel pellets stacked in an axial direction of the sleeve and two sealed closure plugs having a cylindrical part inserted coaxially into an axial end part of the sleeve, the plug welded in a welding station by melting the sleeve and the plug along a circular line in a joint plane perpendicular to the axial direction of the sleeve and the plug through welding directed radially relative to the circular line in the joint plane of the sleeve and the plug, which are coupled to a rotation arrangement for rotation about a common axis, inspection effected by processing digitized optical images of areas of the fuel element adjoining the circular line and distributed along a periphery of the fuel element, the method comprising: rotating the sleeve and the plug about the common axis with aid of the rotation arrangement before welding the plug and the sleeve, the plug and the sleeve configured in a welding position at the welding station, wherein images are taken along the periphery of the fuel element to obtain the digitized optical images which are analyzed to determine a position of the joint plane, and a rotation of the fuel element is verified; deducing at least one of to perform the welding or not to perform the welding; and taking images, if the welding is performed after welding the plug to the sleeve of the fuel element of the fuel in the position at the welding station, along a circumference of the external surface of the fuel element in a vicinity of the line to obtain the digitized optical images, the digitized optical images analyzed to check a conformance of as weld a long the line. 2. The method according to claim 1 , further comprising: claim 1 determining a reference line in the digitized optical images prior to welding, the reference line positioned in a vertical direction of an image corresponding to a circumferential direction of the fuel element, in a plane perpendicular to the axis of the fuel element, and determining grey levels of pixels of the images along search lines perpendicular to the reference line to determine a position of two edges of the joint plane in two planes perpendicular to the axis of the fuel element. 3. The method according to claim 2 , further comprising: claim 2 determining a mean search fine from N adjoining search lines that are determined on a diagram of the grey levels on the mean search line and a detection threshold of the joint plane and comparing the grey levels along the mean search line to the detection threshold to determine the position of the edges of the joint plane, wherein N is an integer greater than 0 determined by a user. 4. The method according to claim 3 , further comprising: claim 3 determining a distance in a direction of the axis of the fuel element between the reference line and at least one edge of the joint plane. 5. The method according to claim 2 , wherein positions of the edges of a weld are determined relative to the reference line. claim 2 6. The method according to claim 1 , wherein during a pulsed laser welding, establishing diagrams of grey levels along each column of the image corresponding to peripheral directions of the fuel element, in planes perpendicular to the axis of the fuel element, and, on a curve obtained for each of the columns of the image, determining transitions between minima and maxima of the curve which have a period consistent with a pulsed laser beam period, and, for each of the columns of the image, determining a number of the transitions having a period equal with the pulsed laser beam period to obtain a curve of a distribution of the transitions along the columns of the image, the curve giving the number of transitions per column which is filtered to obtain a curve of the distribution of the transitions in a vicinity of each edge of the weld, and determining a threshold value according to each edge area of the weld, and determining a condition of edges of the weld by determining columns in which a number of transitions are below the threshold. claim 1 7. The method according to claim 6 , further comprising: claim 6 determining a width of the weld between the edges in the direction of the axis of the fuel element and deducing a depth of penetration of the weld therefrom by correlation. 8. The inspection method according to claim 1 , wherein during a laser beam welding where a fixed reference line is disposed along a laser beam welding axis, determining the position of the joint plane relative to the fixed reference line. claim 1 |
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description | This is a Continuation-In-Part Application of application Ser. No. 15/807,049 filed Nov. 8, 2017, entitled FLOATING NUCLEAR POWER REACTOR WITH A SELF-COOLING CONTAINMENT STRUCTURE AND AN EMERGENCY HEAT EXCHANGE SYSTEM. This invention relates to a floating nuclear power reactor. More particularly this invention relates to a floating nuclear power reactor including a barge which is floatably positioned in the interior of a large water-filled tank or body of water and wherein the nuclear power reactor is positioned on the barge. Even more particularly, the invention relates to a protection system for a floating nuclear power reactor to protect the nuclear reactor from an aircraft strike or a missile strike. Additionally, the protection system of this invention includes structure to reduce the impact forces of an aircraft strike or a missile strike. Applicant has received U.S. Pat. Nos. 9,378,855; 9,396,823; and 9,502,143 relating to nuclear reactors positioned in a body of water to be able to flood and cool the nuclear reactor in the event of overheating or over pressurization of the nuclear reactor. In Applicant's latest invention shown and described in the co-pending application Ser. No. 15/807,049 filed Nov. 8, 2017, a suspension system is described for suspending and stabilizing a barge which is floating in a large water tank. That system is incorporated herein which further enhances the protection of the nuclear reactor in the event of an aircraft strike or a missile strike. In Applicant's co-pending patent application Ser. No. 15/807,049 filed Nov. 8, 2017, the upper end of the barge, having the nuclear reactor mounted thereon, is exposed to an aircraft strike or a missile strike. The instant invention is to provide a protection system for the nuclear power reactor of the co-pending application and to provide protection to other exposed nuclear power reactors of different designs. This Summary is provided to introduce a selection of concepts in a simplified form that are further described below in the Detailed Description. This Summary is not intended to identify key aspects or essential aspects of the claimed subject matter. Moreover, this Summary is not intended for use as an aid in determining the scope of the claimed subject matter. A floating nuclear reactor is disclosed. The floating nuclear reactor of this invention includes a tank, which may be rectangular, having a bottom wall, an upstanding first end wall, an upstanding second end wall, an upstanding first side wall and an upstanding second side wall. Each of the first end wall, the second end wall, the first side wall and the second side wall of the tank have an outer side, an inner side, a lower end and an upper end. The tank is partially buried in the ground with the tank having water therein. A barge is floatably positioned in the tank with the barge having a bottom wall, a first end wall, a first side wall, a second side wall and an open second end. A nuclear reactor is positioned on the barge. At least one suspension assembly, and preferably two suspension assemblies, connect the first end wall of the barge to the first end wall of the tank. At least one suspension assembly, and preferably two suspension assemblies, connect the first side wall of the tank to the first side wall of the barge. At least one suspension assembly, and preferably two suspension assemblies, connect the second side wall of the tank to the second side wall of the barge. At least one suspension assembly, and preferably two suspension assemblies, connect the second end wall of the barge to the second end wall of the tank. The suspension assemblies permit the barge to move upwardly and downwardly with respect to the tank while maintaining the barge in a level condition. The suspension assemblies permit the barge to move downwardly if struck by a missile or aircraft to lessen the impact thereof. The nuclear reactor is positioned in the tank so as to close the open second end of the barge. The nuclear reactor may also be positioned in a body of water. The nuclear reactor includes a first containment member which has a cylindrical body portion, a hemi-spherical upper end and a hemi-spherical lower end. The first containment member is comprised of stainless steel or other suitable material. The first containment member is positioned at the open end of the barge with the sides of the containment member being in engagement with the ends of the sidewalls of the barge so as to close the open end of the barge. The positioning of the first containment member causes the outer side of the first containment member to be in contact with the water in the tank. The first containment member defines a sealed interior compartment. The first containment member has a hatch or door mounted thereon at the lower end thereof which selectively closes an opening in the first containment member. The first containment member also has a pipe extending from the lower end thereof which is in fluid communication with the interior compartment thereof. A normally closed one-way valve is imposed in the pipe. A reactor vessel is positioned in the interior compartment of the first containment member. The nuclear reactor of the co-pending application has a unique cooling system for the nuclear reactor which does not form a part of this invention. A heat exchanger is positioned adjacent the first containment member and includes a body section, an upper section and a lower section. The heat exchanger includes an outer wall member or second containment member which is comprised of metal. A vessel is positioned within the second containment member of the heat exchanger. The vessel has an interior compartment which is filled with fluid. The heat exchanger is connected to a turbine or other device. The suspension assemblies also permit the barge to move downwardly in the tank in the event of an aircraft strike, a missile strike or an earthquake to reduce the impact forces on the barge and nuclear reactor. A hollow steel conical-shaped member is mounted on the upper end of the first containment member of the nuclear reactor. A hollow steel conical-shaped member is also mounted on the upper end of the heat exchanger. If an aircraft or a missile should strike either of the conical-shaped members, the conical-shaped members would cause the disintegration of the aircraft or missile and would deflect the same. A roof is positioned over the upper end of the barge which hides the location of the nuclear reactor and heat exchanger from view so that an aircraft attempting to strike either the nuclear reactor or the heat exchanger will not know the precise position of those structures on the barge. In the preferred embodiment of the invention, the interior of the conical-shaped members on the upper ends of the nuclear reactor and the heat exchanger will be filled with a material which acts as an impact absorbing member. It is therefore a principal object of the invention to provide a floating nuclear reactor protection system. A further object of the invention is to provide a conical-shaped member on the upper end of the nuclear reactor which will cause an aircraft or missile to disintegrate upon impacting the conical-shaped member; Yet another object of the invention is to provide a protection system for a floating nuclear reactor which includes a roof over the protection system. These and other objects will be apparent to those skilled in the art. Embodiments are described more fully below with reference to the accompanying figures, which form a part hereof and show, by way of illustration, specific exemplary embodiments. These embodiments are disclosed in sufficient detail to enable those skilled in the art to practice the invention. However, embodiments may be implemented in many different forms and should not be construed as being limited to the embodiments set forth herein. The following detailed description is, therefore, not to be taken in a limiting sense in that the scope of the present invention is defined only by the appended claims. Applicant has previously received U.S. Pat. Nos. 9,378,855; 9,396,823; and 9,502,143 relating to floating nuclear power reactors. Applicant incorporates the disclosure of the above identified patents in their entirety by reference thereto to complete this disclosure if necessary. The floating nuclear reactor of this invention is referred to generally by the reference numeral 10. The nuclear reactor 10 floats in a concrete tank 12 having a bottom wall 14, a first end wall 16, a second end wall 18, a first side wall 20, a second side wall 22 and an open upper end 24. Tank 12 is buried in the ground 26 as seen in FIG. 1 so that the open upper end 24 of tank 12 is at or above ground level 28. The tank 12 is partially filled with water 30 from a source of water. Preferably the water 30 is gravity fed to the tank 12. The tank 12 may be completely buried in the ground. The numeral 32 refers to a barge-like vessel which floats in the tank 12. Barge 32 includes a bottom wall 34, a first side wall 36, a second side wall 38, a semi-circular end wall 40 and an open end 41 at the ends 42 and 43 of side walls 36 and 38 respectively. Barge 32 is comprised of a metal material such as stainless steel, steel, iron, aluminum or other suitable material. Barge 32 is supported in tank 12 by a plurality of upper suspension assemblies 44, 46, 48, 50, 52, 54, 56 and 58 which extend between the barge 32 and the tank 12 as will be described in detail hereinafter. Barge 32 is also supported in tank 12 by eight lower suspension assemblies, identical to suspension assemblies 44, 46, 48, 50, 52, 54, 56 and 58, which are positioned below suspension assemblies 44, 46, 48, 50, 52, 54, 56 and 58. The numeral 59 refers to a nuclear reactor which is positioned in barge 32 so as to close the open end 41 of barge 32 as will be explained in detail hereinafter. Reactor 59 includes an upstanding containment member 60 which has a cylindrical body portion 62, a hemi-spherical upper end 64 and a hemi-spherical lower end 66. Containment member 60 is comprised of stainless steel or other suitable material. Containment member 60 is positioned at the open end 41 of barge 32 with the sides of containment member 60 being in engagement with the ends 42 and 43 of side walls 36 and 38 respectively of barge 32 and being secured thereto by welding or the like to close the open end 41 of barge 32. The positioning of the containment member 60 as just described causes the outer side of containment member 60 to be in contact with the water 30 in tank 12. Containment member 60 defines a sealed interior compartment 68. Containment member 60 has a hatch 70 mounted therein as seen in FIG. 3. Containment member 60 also has a pipe 72 extending from the lower end thereof which is in fluid communication with the interior compartment 68. A normally closed one-way valve 74 is imposed in pipe 72. A reactor vessel 75 is positioned in compartment 68 and has an interior compartment 76. Vessel 75 is supported in compartment 68 by braces 77 which extend between the exterior of reactor vessel 75 and the interior side of containment member 60 as seen in FIG. 3. The numeral 80 refers to an upstanding heat exchanger which is positioned adjacent containment member 60 as seen in the drawings. Heat exchanger 80 includes a body section 82, an upper section 84 and a lower section 86. Heat exchanger 80 is comprised of a metal material such as stainless steel or other suitable material. A vessel 88 is positioned within heat exchanger 80 and is supported therein by braces 90 extending therebetween. Vessel 88 defines an interior compartment 92. A tube 94 interconnects the reactor vessel 75 and the vessel 88 of heat exchanger 80 as seen in the drawings. The heat exchanger 80 is connected to a turbine 96 or other device which is connected to a generator 98 or other structure. A hollow metal cone 100 is mounted on the hemi-spherical upper end 64 of containment member 60. Cone 100 is comprised of stainless steel, steel or other suitable material. Cone 100 has an interior compartment 102 which is preferably filled with a filter material 104 which not only may serve as a filtration bed but serves as an impact absorber should the cone 100 be struck by an aircraft or a missile. The cone 100, if struck by an aircraft or missile, will disintegrate or tear apart the aircraft or missile and deflect the aircraft or missile away from the cone 100. An outlet pipe 106 may be provided in the upper end of containment member 60 to permit steam or the like to pass upwardly therethrough onto the filtration material 104. The cone 100 may also have a discharge tube assembly 108 extending upwardly from pipe 106 and which has discharge tubes 110 extending downwardly and outwardly from the upper end of tube 108. A metal cone 112 extends upwardly from the upper end of heat exchanger 80 and is filled with an impact absorbing material 114. Cone 112, if struck by an aircraft or missile, will disintegrate the aircraft or missile in the same manner as the cone 100. A roof 116 extends over the cones 100, 112 and the barge 32 to hide the reactor 59 and the heat exchanger 80 from view. Thus, if an aircraft is attempting to strike the reactor 59, the pilot of the aircraft will not be able to determine the exact location of the reactor 59. A pair of vertically disposed guide tracks or channels 120 and 122 are secured to the inner side of end wall 18. A pair of vertically disposed guide tracks or channels 124 and 126 are secured to the inner side of side wall 20. A pair of vertically disposed guide tracks or channels 128 and 130 are secured to the inner side of end wall 16. A pair of vertically disposed guide tracks or channels 132 and 134 are secured to the inner side of side wall 22. Each of the guide tracks 120, 122, 124, 126, 128, 130, 132 and 134 have an upper wheel and a lower wheel vertically movable therein. The guide tracks 134, 120, 122, 124, 126, 128, 130 and 132 form a part of the suspension assemblies 44, 46, 48, 50, 52, 54, 56 and 58 respectively. In as much as the suspension assemblies 44, 46, 48, 50, 52, 54, 56 and 58 are identical except for length, only suspension assembly 48 will be described in detail. Suspension assembly 48 includes an upper chain member 136, a lower chain member 138 and an intermediate chain member 140. The outer ends of chain members 136, 138 and 140 are secured to the upper wheel in guide track 122. The inner ends of chain members 136, 138 and 140 are secured to the barge 32. As seen, upper chain member 136 extends upwardly and inwardly from guide track 122 to barge 32. As also seen, lower chain member 138 extends downwardly and inwardly from guide track 122 to barge 32. Further, as seen, intermediate chain member 140 extends horizontally inwardly from guide track 122 to barge 32. The suspension assembly below suspension assembly 46 would be similarly attached to the lower wheel in guide track 122 and the barge 32. The other suspension assemblies would be attached to the guide tracks 124, 126, 128, 130, 132 and 134 and the barge 32. The suspension assemblies 44, 50, 56 and 58 are identical. The suspension assemblies 46, 48, 54 and 56 are identical. The only difference between the suspension assemblies 44, 50, 56, 58 and the suspension assemblies 46, 48, 54 and 56 is that the suspension assemblies 46, 48, 54 and 56 are somewhat longer than the suspension assemblies 44, 50, 56 and 58. As stated in the co-pending patent application, the guide tracks or channels could be secured to the barge rather than being secured to the walls of the tank. In that embodiment, the ends of the chains of the suspension assemblies would be secured to the tank. Although it is preferred that each of the suspension assemblies have a horizontally disposed intermediate chain member 140, the intermediate chain member 140 may be omitted in some situations. If an aircraft or a missile should strike either of the cones 100 or 112, the cones will disintegrate and deflect the aircraft or the missile to prevent damage to the nuclear reactor. Additionally, the impact absorbing material in the cones 100 and 112 will lessen the damage to the nuclear reactor. Further, if the barge 32 or the cones 100 and 112 are struck by an aircraft or missile, the suspension systems will permit the barge 32 to move downwardly in the tank 12 to lessen or absorb the impact forces of the strike. Thus it can be seen that the invention accomplishes at least all of its stated objectives. Although the invention has been described in language that is specific to certain structures and methodological steps, it is to be understood that the invention defined in the appended claims is not necessarily limited to the specific structures and/or steps described. Rather, the specific aspects and steps are described as forms of implementing the claimed invention. Since many embodiments of the invention can be practiced without departing from the spirit and scope of the invention, the invention resides in the claims hereinafter appended. |
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abstract | A repair apparatus for a shroud in a nuclear reactor pressure vessel is described. The repair apparatus includes an upper stabilizer assembly, a lower stabilizer assembly, and a tie rod configured to extend between and to couple to the upper and lower stabilizer assemblies. The upper stabilizer includes a stabilizer block and an upper stabilizer wedge slidably coupled to the upper stabilizer block. The upper stabilizer block is configured to couple to a shroud lug. The upper stabilizer wedge includes an integral leaf spring portion formed by a slot in the wedge and is configured to engage the side wall of the reactor pressure vessel. The lower stabilizer assembly includes a stabilizer block and a lower stabilizer wedge slidably coupled to the lower stabilizer block. The lower stabilizer block is configured to engage the shroud. A horizontal stabilizing spring is attached to the wedge and is configured to engage the side wall of the reactor pressure vessel. The tie rod is threaded at each end. One end threadedly engages a tie rod opening in the bottom stabilizer block. The other end is received by the upper stabilizer block and is secured by a tie rod nut. |
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description | An exemplary embodiment of the present invention may include state sensing for energy stores, such as automobile batteries, but is not limited to this application. FIG. 1 shows the components for state sensing according to an exemplary embodiment of the present invention of an energy store 1, such as an automobile battery. A sensor and measurement unit 2 perform measurements of performance quantities x, such as current, voltage, and/or temperature, on battery 1. The measured performance quantities are supplied by lines 7 to a state estimator 3, which, for example, may determine state variables that characterize the current system state using a Kalman filter. Such state variables may include the available charge or the age of battery 1. State estimator utilizes a model, in which measured performance quantities x are entered to determine the state variables a. The model operates using model parameters p, which are also dependent on the aging processes of energy store 1. To avoid the model losing its validity due to changed parameters p, model parameters p are updated using a parameter estimator 4. For this purpose, a parameter estimation routine is used, which uses measured performance quantities x and may also use additionally estimated state variables a as input quantities. Updated parameters p are then delivered to state estimator 3. For this purpose, state estimator 3 and parameter estimator 4 are connected to one another. State variables a, determined by state estimator 3, are processed further to perform favorable measures (for example, charge state displays, modification of the energy supply). FIG. 1B shows a suitable state estimator 3 and parameter estimator 4, in which the individual components for state sensing, according to an exemplary embodiment of the present invention, are each present and assembled into one unit. Measured performance quantities x are supplied by lines 7 to state estimator 3 and/or parameter estimator 4. Subtractors or differentiators, which produce gradients of one measurement quantity x at a time, are used as a detection arrangement 8 for detecting the dynamic response of measured performance quantities x. A selection unit 9, which selects state variable a and/or parameters p subsequently estimated depending on the detected dynamic response of the performance quantities x, is connected downstream. Selected performance quantities x are supplied at state estimator 3, together with updated parameters p, to a computation unit 10, which computes specific state variables a using a model. Most estimation models operate with covariance matrices, the values assigned to the individual state variables of which converge toward zero, if the estimated value approximates the real value over time. These matrix values (covariances) may therefore be used for rating the quality of the estimation. To rate the quality of the estimation, threshold values associated with the respective covariances may be, for example, fixed in a unit 11, and the quality of the estimation may be determined by subtracting the estimated value from the fixed threshold value. If, for example, an estimated state variable does not fall below the threshold value after a predetermined number of cycles, the estimated value may be rejected and the previously estimated value may be maintained instead. In this way, increasing deterioration of the estimation may be prevented. FIG. 2A shows an example of a rapidly converging estimated state variable a(3), which is not subject to any fluctuations after convergence. Such state variables, such as the concentration overvoltage, have large time constants. The associated matrix element of the covariance matrix shown in FIG. 2B, in this example K(3,3) to a(3), i.e., the covariance to this state variable, converges rapidly toward zero. To check the quality of the estimation, a threshold that may be reached after a certain number of cycles may be fixed, i.e., a number of iterative estimations. If not, the estimation for the state variable may be rejected. An example of a divergence of a current state variable xc3xa3(1) and associated estimated value a(1) is shown in FIG. 3. The fluctuating time curve of current state variable xc3xa3(1) and estimated state value a(1), which moves away from the zero line over time, is shown in FIG. 3A below the zero line. Associated covariance K(1,1) to state variable a(1) indicates that the estimation may not be suitable. The covariance does not converge, but increases continuously over time, as shown in FIG. 3B. Cases, such as that of FIG. 3, may be avoided by an exemplary embodiment of the present invention, which utilizes xe2x80x9cbackupxe2x80x9d methods, if the quality of the estimation is not sufficient. FIG. 4 shows a flow diagram of an exemplary method according to the present invention. At the beginning of the estimation method, a specific time Tmin1 passes, before the system assumes a state suitable for state estimation, which occurs before the actual estimation method begins. Subsequently, the dynamic response of the excitation, i.e., the dynamic response of measured performance quantities x, is scanned (S1). These may be, for example, time-dependent quantities current, temperature, and voltage. If, for example, the discharge current of the battery remains nearly zero over a relatively long period of time, since, for example, the consumer may be completely supplied by the generator, specific state variables a or parameters p dependent on the current may not be subject to change. Further measurement values are then awaited, until a further time interval Tmin2 passes (S2). If a dynamic response of the measured performance quantities begins, the quantity of the dynamic response is scanned (S3). For a low dynamic response of the measurement values, it is first determined whether the system is in a limit state or boundary region (in batteries, for example, the fully charged or drained state). The same scanning also occurs if there is a large dynamic response of the measured performance quantities (S4 and/or S5). If the system is not in a boundary region, the actual estimation of the state variables may be started. According to an exemplary embodiment of the present invention, at a low dynamic response of the measurement values, state variables having small time constants are maintained (S6), while state variables having large time constants are estimated (S7). In contrast, for measurement values having a large dynamic response, the state variables having large time constants are maintained (S8), while the state variables having small time constants are estimated (S9). In the battery application described above, the ohmic values and the charge-transfer overvoltage represent state variables/parameters having small time constants, while, for example, the concentration overvoltage may represent a state variable having a large time constant. According to an exemplary embodiment of the present invention, the parameters and state variables that are not expected to cause changes in the dynamic response of the system are not redetermined by estimation. In this way, enlarging inaccuracies during the estimation due to unnecessarily frequent estimations, which may invalidate the model or provide incorrect state results, may be avoided. If, while checking (S4, S5) whether the system is in a limit state (boundary region), it is determined that the system is in a limit state (boundary region), the state variables/parameters may be, for example, maintained or evaluated using xe2x80x9cbackupxe2x80x9d methods (S10, S11), to avoid incorrect estimations (boundary regions of the model accuracy). These methods are based on stable models, in which no divergence is expected. After the routine shown in FIG. 4 is finished, one cycle of the state estimation, according to an exemplary embodiment of the present invention, ends, and further cycles may follow immediately or with delays, which may be fixed. FIG. 5 shows a flow diagram for determining the quality of the estimation described above. For this purpose, a measurable quantity, which is calculated from estimated quantities, is compared with the quantity actually measured (T1). If there is good correspondence (for example of the estimated and measured battery voltages), the convergence of the covariances associated with the state variables/parameters is checked (T2). Specifically, individual covariances may not yet have sufficiently converged (see also FIG. 2B), so that a specific typical convergence time Tmin still should be awaited (T3) until sufficiently good convergence results. When this occurs, the estimated state variables/parameters are evaluated (T4) and from this, for example, the charge state or the age of the battery may be determined. In contrast, if time Tmin has already passed without the associated covariances having sufficiently converged, i.e., for example, having passed below a specific threshold value, the estimated quantities are rejected and the parameter estimation routine and/or the state estimation routine (Kalman filter) is restarted (T5). Until the reestimated quantities are received, simple xe2x80x9cbackupxe2x80x9d methods may be utilized (T6). If there is no sufficient correspondence between an easily measurable and estimable reference quantity (for example battery voltage) from the start, the covariance matrix may not be sufficiently converged. The result may be rechecked after a time period Tmin* (T7). If the result remains unchanged, the parameter and/or state estimation is restarted (see FIG. 4) and xe2x80x9cbackupxe2x80x9d methods may be utilized (T8, T9). |
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description | This application claims priority to Korean Patent Application No. 10-2008-46748 filed on May 20, 2008, under 35 U.S.C. §119, the entire contents of which are incorporated herein by reference. 1. Field Example embodiments relate to an electron beam focusing electrode and an electron gun using the same. Particularly, example embodiments relate to an electron beam focusing electrode that reduces a spreading phenomenon of electron beams by passing electron beams radiated from a cathode electrode of the electron gun through a through-hole having a desired and/or predetermined sectional shape, as well as an electron gun including the electron beam focusing electrode. 2. Discussion of the Related Art In manufacturing a vacuum device for oscillation of microwaves and terahertz waves, an electron gun is used for allowing electron beams to be irradiated onto the device. A conventional electron gun generates an electron beam having a solid or annular section. In order to utilize an electron beam having a solid or annular section, the electron beam should be incident into a pattern formed on a surface of a substrate, or the like. However, as the size of a device becomes smaller and smaller, it is more and more difficult to allow an electron beam to be incident into a fine pattern. Another conventional electron gun generates an electron beam having a rectangular section. However, the electron beam having a rectangular section generated by the conventional electron gun has less laminarity than a solid or annular beam. The above information disclosed in this Background section is only for enhancement of understanding of the background of the present application. Therefore, it may contain information that does not form the prior art that is already known in this country to a person of ordinary skill in the art. Example embodiments are provided at least in part to address issues, which may prevent conventional devices from outputting a predetermined and/or desired beam. For example, there is provided device and a method to address an issue relating to less laminarity than a solid or annular beam. An example embodiment provides an electron beam focusing electrode, which may be included in an electron gun. The electron beam focusing electrode may include a plate having a polygonal through-hole and a projecting portion formed on at least one side of the through-hole. According to an example embodiment, the projecting portion may be spaced apart from both ends of the side on which the projecting portion is formed. A length of the projecting portion may be smaller than the distance from a center of the through-hole to the side on which the projecting portion is formed. According to an example embodiment, an inner surface of the through-hole is inclined with respect to a traveling direction of an electron beam passing through the through-hole. The through-hole may have a first area and a second area. The first area may be smaller than the second area. Further, the first area may be an incident area of an electron beam, and the second area may be an emission area of the electron beam. According to an example embodiment, the polygonal through-hole may include four sides, and four projecting portions respectively arranged on the four sides. Each projecting portion may protrude from a center of the respective side. Each projecting portion may have a rectangular cross section. Another example embodiment provides an electron gun. The electron gun may include an electron beam focusing electrode such as the electron beam focusing electron described above in this summary. The electron gun may also include a cathode electrode radiating electrons and an anode electrode spaced apart from the cathode electrode and on which the electrons radiated from the cathode electrode are focused. According to an example embodiment, the electron beam focusing electrode of the electron gun may be electrically isolated from the cathode electrode of the electrode gun. Alternatively, the electron beam focusing electrode of the electron gun may be connected to the cathode electrode of the electron gun. According to an example embodiment, the electron gun may include a gate electrode positioned between the electron beam focusing electrode and the anode electrode to adjust a current quantity of an electron beam. According to an example embodiment, the cathode electrode of the electron gun may be one of a cold emission cathode, a photocathode and a plasma source. The electron gun may also include a heat shield mounted around the cathode electrode to shield heat radiated from the cathode electrode. Still another example embodiment provides a method of reducing a spreading phenomenon of an electron beam with rectangular cross section. The method may include forming an electric field in a polygonal through-hole having a projection portion arranged on at least one side of the through-hole, passing an electron beam through the through-hole, and forming a predetermined cross section for the electron beam by the electric field. The method may also include using a gate electrode to adjust a current quantity of the electron beam. It should be understood that the appended drawings are not necessarily to scale, present a somewhat simplified representation of various preferred features illustrative of the basic principles of this disclosure. The specific design features disclosed herein, including, for example, specific dimensions, orientations, locations and shapes will be determined in part by the particular intended application and use environment. In the figures, like reference numerals refer to the same or equivalent parts of example embodiments throughout the following detailed description. Example embodiments are described more fully hereinafter with reference to the accompanying drawings, in which example embodiments are shown. Example embodiments may, however, be embodied in many different forms and should not be construed as limited to the example embodiments set forth herein. Rather, these embodiments are provided so that this disclosure will be thorough and complete, and will fully convey the scope of example embodiments to those skilled in the art. In the drawings, the size and relative sizes of regions may be exaggerated for clarity. It will be understood that when an element is referred to as being “on,” “connected to” or “coupled to” another element and the like, it can be directly on, connected or coupled to the other element or layer or intervening elements or layers may be present. In contrast, when an element is referred to as being “directly on,” “directly connected to” or “directly coupled to” another element or layer, there are no intervening elements present. 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, although the terms first, second, third etc. may be used herein to describe various elements, components, regions, and/or sections, these elements, components, regions and/or sections should not be limited by these terms. These terms are only used to distinguish one element, component, region or section from another region or section. Thus, a first element, component, region or section discussed below could be termed a second element, component, region or section without departing from the teachings of example embodiments. Spatially relative terms, such as “beneath,” “below,” “lower,” “above,” “upper” and the like, may be used herein for ease of description to describe one element or feature's relationship to another element(s) or feature(s) as illustrated in the figures. It will be understood that the spatially relative terms are intended to encompass different orientations of the device in use or operation in addition to the orientation depicted in the figures. For example, if the device in the figures is turned over, elements described as “below” or “beneath” other elements or features would then be oriented “above” the other elements or features. Thus, the exemplary term “below” can encompass both an orientation of above and below. The device may be otherwise oriented (rotated 90 degrees or at other orientations) and the spatially relative descriptors used herein interpreted accordingly. 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 context clearly indicates otherwise. It will be further understood that the terms “comprises” and/or “comprising,” when used in this specification, 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. FIG. 1A is a cross-sectional perspective view of an electron gun according to example embodiments, and FIG. 1B is a longitudinal cross-sectional view of the electron gun shown in FIG. 1A. Referring to FIG. 1A, the electron gun may include a cathode electrode 10, an anode electrode 20 and an electron beam focusing electrode 30. The cathode electrode 10 may be a device to radiate electrons. For example, the cathode electrode 10 may be a device using thermionic emission, or may be a cold emission cathode, a photocathode or a plasma source. Referring to FIG. 1B, the cathode electrode 10 may be fixed at a desired and/or predetermined position in the electron gun by a cathode base 100 and a cathode support sleeve 101 according to an example embodiment. If the cathode electrode 10 is a device using thermionic emission, a heat shield 102 for shielding heat radiated from the heated cathode electrode 10 may be mounted around the cathode electrode 10. The anode electrode 20 may be spaced apart from the cathode electrode 10 at a desired and/or predetermined distance. A voltage may be applied between the cathode electrode 10 and the anode electrode 20. Electrons radiated from the cathode electrode 10 may be accelerated by the applied voltage, so that electron beams may be formed in a direction towards the anode electrode 20. Further, the anode electrode 20 may have a hole 21 at the center thereof, according to an example embodiment. Electrons radiated from the cathode electrode 10 may pass the anode electrode 20 through the hole 21 to be emitted from the electron gun and may reach a collector (not shown) thereafter. The collector may be an anode electrode positioned outside the electron gun. Referring to FIG. 1B, the electron beam focusing electrode 30 may be fixed at a desired and/or predetermined position between the cathode electrode 10 and the anode electrode 20 by a cylinder-shaped base 300. In an example embodiment, the electron beam focusing electrode 30 may include a plate having a polygonal through-hole 33 formed therein, so that a more desirable electric field may be formed. When electrons radiated from the cathode electrode 10 pass the electron beam focusing electrode 30 through the through-hole 33, an electron beam may be formed to have a predetermined and/or desired cross-sectional shape. According to an example embodiment, the electron gun may further include a gate electrode (not shown) positioned between the electron beam focusing electrode 30 and the anode electrode 20 for adjusting the current quantity of an electron beam. FIG. 2 is an enlarged longitudinal cross-sectional view showing a vicinity of the through-hole 33 of the electron gun, according to example embodiments. Referring to FIG. 2, the electron beam focusing electrode 30 may be positioned in front of the cathode electrode 10 from which electrons may be radiated. The cathode electrode 10 may be surrounded by the cathode sleeve 12. The cathode sleeve 12 may have a desired and/or predetermined emission hole 11. Electrons radiated from the cathode electrode 10 may be emitted in a direction toward the electron beam focusing electrode 30 through the emission hole 11 of the cathode sleeve 12. When the emitted electrons pass through the through-hole 33 of the electron beam focusing electrode 30, an electron beam may be formed and a sectional shape of the electron beam may be determined by an electric field formed therein. The electric field may be formed depending on the shape of the through-hole 33. The cathode electrode 10 and the cathode sleeve 12 will be described later with reference to FIG. 4. Referring to FIG. 2, the cathode sleeve 12 and the electron beam focusing electrode 30 may be spaced apart from each other at a desired and/or predetermined distance and may be electrically isolated from each other, according to an example embodiment. Therefore, the electron beam focusing electrode 30 may be electrically isolated from the cathode electrode 10, which may be connected to the cathode sleeve 12. According to an example embodiment, the cathode electrode 10 and the electron beam focusing electrode 30 may have the same electric potential or may have different electric potentials to control a trace of the electron beam. When different electric potentials are applied to the cathode electrode 10 and the electron beam focusing electrode 30, a potential difference between the cathode electrode 10 and the electron beam focusing electrode 30 may be determined that does not breakdown the isolation between the cathode electrode 10 and the electron beam focusing electrode 30. According to another example embodiment, the electron beam focusing electrode 30 and the cathode electrode 10 may be connected to each other. For example, the electron beam focusing electrode 30 and the cathode electrode 10 may be connected through the cathode sleeve 12 by connecting the electron beam focusing electrode 30 to the cathode sleeve 12. FIG. 3A is a perspective view of an electron beam focusing electrode according to an example embodiment. Referring to FIG. 3A, the electron beam focusing electrode 30 may include a plate 30′ having a first surface 31, a second surface opposing the first surface 31, and a polygonal through-hole 33 passing through the electron beam focusing electrode 30. The polygonal through-hole 33 may include projecting portions 34 protruding inside the through-hole 33 from respective sides of the through-hole 33. For example, the polygon of the through-hole 33 may have four sides. Each side may have one projecting portion formed on the center of the side. Each projecting portion may have a rectangular cross-section and may protrude from each side of the polygon. Electrons radiated from a cathode electrode may be incident onto the first surface 31 of the electron beam focusing electrode 30. Because the through-hole 33 may be formed to pass through the first surface 31 and the second surface 32, the electrons may incident to the through-hole 33 from the first surface 31, pass through the through-hole 33, and then may be emitted from the through-hole 33 from the second surface 32. Referring to FIG. 3A, the through-hole 33 may further include at least one projecting portion 34 formed on at least one side of the through-hole 33. Distortion of an electric field at an edge of the electron beam may be reduced due to the projecting portion 34 and traces of electrons passing through the through-hole 33 may be controlled. Consequently, the laminarity of electron beams emitted from the electron gun may be improved. FIG. 3B is a plan view of the second surface 32 of the electron beam focusing electrode, and FIG. 3C is a bottom view of the first surface 31 of the electron beam focusing electrode, according to an example embodiment. In the electron beam focusing electrode 30, the through-hole 33 of the plate 30′ may have a first sectional area at the first surface 31 shown in FIG. 3C and a second sectional area at the second surface 32 shown in FIG. 3B. The first sectional area may be different from the second sectional area. For example, the second sectional area may be larger than the first sectional area. As a result, the section of the through-hole 33 may be formed to be inclined with respect to the traveling direction of an electron beam passing through the through-hole 33. As shown in FIG. 3B, the through-hole 33 formed in the plate 30′ at the second surface 32 has a length L1 and a width H1 in lateral and longitudinal directions of FIG. 3B, respectively. For example, a through-hole may have L1=3.04 mm and H1=2 mm. At least one projecting portion 34 may be formed on at least one side of the through-hole 33. Each of the projecting portions 34 may be spaced apart with desired and/or predetermined distances from both ends of the respective side, on which the projecting portion 34 is formed. Each of the projecting portions 34 may be protruded by a desired and/or predetermined height towards a central direction of the through-hole 33. For example, as shown in FIG. 3B, the through-hole 33 may have one projecting portion protruded on each of the sides of the polygon, respectively, towards the center of the through-hole 33, each of the projecting portions 34 may be positioned at the center of the respective side and may be positioned apart from both of the two ends on the left side and the right side of the projection portion. The projecting portions 34 in the lateral and longitudinal directions may have lengths L2 and H2, and lengths D1 and D2, respectively. The length D1 or D2 of each of the projecting portions 34 may be formed to be smaller than the distance between the respective side to the center of the through-hole 33, so that two opposing projecting portions 34 may not protrude to touch each other. For example, projecting portions may have L2=0.88 mm, H2=0.48 mm, and D1=D2=0.4 mm. Accordingly, the rectangular shaped through-hole 33 may be modified into a dumbbell shaped polygon by the projecting portions 34 protruded from each side of the rectangular through-hole 33. Consequently, the electric field in the through-hole 33 may be modified by the dumbbell shape of the through-hole 33, so that a spreading phenomenon of an electron beam at corners of the through-hole 33 may be reduced compared to a through-hole having a rectangular shape or a rectangular shape with curved corners. When an electron beam passes through a rectangular or curved-corner rectangular through-hole of an electron beam focusing electrode, symmetry of electron distribution may be disrupted as a traveling distance of the electron beam increases. This may be because the electron beam is influenced by the distribution of electric field depending on the shape of the electron beam focusing electrode. This may also be because the initial velocity of the spread and initial electron speed due to the non-uniformity of a distribution of heat and electric field at the earlier stage of the electron beam generation. If the aforementioned electron beam focusing electrode 30 with the dumbbell shape through-hole 33 is used, the trace of an electron beam passing through the through-hole may be controlled by the projecting portions 34. Consequently, a uniformity of the electron beam may be improved and/or a more uniform electron beam may be obtained. FIG. 3C is a bottom view showing the first surface 31 of the electron beam focusing electrode 30. According to an example embodiment shown in FIG. 3C, the electron beam focusing electrode 30 may be formed by joining two circular electrodes having different diameters together. Alternatively, the electron beam focusing electrode 30 may also have a shape other than a circular shape or may include a number of electrodes other than two pieces. The through-hole 33 formed in the plate 30′ may have a lengths L3 and H3 in the lateral and longitudinal directions of the first surface 31, respectively. For example, a through-hole may have L3=2.2 mm and H3=1.16 mm. At least one projecting portion 34 may be formed on at least one side of the through-hole 33. Each of the projecting portions 34 may be spaced apart with desired and/or predetermined distances from both ends of the respective side, on which the projecting portion 34 is formed. For example, each side of the through-hole 33 may have a projecting portions 34 formed at a center of the side, protruding to a center of the through-hole 33. The projecting portions 34 may have widths L2 and H2, and lengths D1 and D2, in the lateral and longitudinal directions, respectively. FIGS. 3D and 3E are cross-sectional views of the electron beam focusing electrode shown in FIGS. 3A along A-A and B-B, respectively, according to example embodiments. Referring to FIGS. 3D and 3E, the through-hole 33 formed in the plate 30′ may be formed such that the sectional area at the second surface 32 of the electron beam focusing electrode 30 is larger than that at the first surface 31 of the electron beam focusing electrode 30. As a result, an inner surface 331 of the through-hole 33 may have an angle of θ with respect to the first surface 31. Further, the through-hole 33 may have a thickness T1. For example, an electron beam focusing electrode may have a through-hole in which θ=50 degrees and T1=0.5 mm. FIG. 4A is an enlarged perspective view showing a portion of a cathode electrode 10 included in an electron gun according to an example embodiment, and FIG. 4B is a longitudinal cross-sectional view of the cathode electrode 10 shown in FIG. 4A. Referring to FIGS. 4A and 4B, the cathode electrode 10 may be positioned in a cathode sleeve 12 having an emission hole 11 formed therein. Electrons radiated from the cathode electrode 10 may be emitted towards the electron beam focusing electrode 30 through the emission hole 11. An inner surface 111 of the emission hole 11 may be formed to make an angle of δ with respect to a surface 110 of the cathode electrode 10. Further, the emission hole 11 may have a thickness T3. For example, an emission hole may have δ=30 degrees and T3=0.06 mm. FIG. 4C is a plan view showing the emission hole 11 shown in FIGS. 4A and 4B. Referring to FIG. 4C, the emission hole 11 may be formed to have a rectangular section having lengths L4 and H4 in the lateral and longitudinal directions of FIG. 4C, respectively. For example, a cathode sleeve may have an emission hole in which L4=0.6 mm and H4=0.1 mm. Referring back to FIG. 2, the electron beam focusing electrode 30 may be positioned to connect with or be spaced apart at a predetermined and/or desired distance from the cathode sleeve 12. Electrons may be radiated from the cathode electrode 10, and then may be emitted through the emission hole 11 to form an electron beam. A predetermined and/or desired sectional shape of the electron beam may be formed by an electric field when the electron beam passes through the electron beam focusing electrode 30. FIG. 5 is a schematic view showing equipotential lines and traces of electrons in an electron beam that passes through an electron beam focusing electrode according to example embodiments. As shown in FIG. 5, the equipotential lines of the electron beam focusing electrode are controlled under the influence of projecting portions protruded from respective sides of the through-hole. As shown in FIG. 5, if the electron beam focusing electrode having projecting portions protruded inside a through-hole is used, distortion of an electron beam distribution may be improved at corners of the electron beam. As a result, distortion and crossing at corners of an electron beam may be decreased and/or prevented and a shape of the electron beam cross section may not change significantly with respect to the distance that the electrons travel. Therefore, the shape of the electron beam cross section may be sustained longer. Although example embodiments have been particularly shown and described with reference to FIGS. 1A-5, those skilled in the art will understand that various modifications, additions and substitutions in forms and details may be made without departing from the scope and spirit of example embodiments. |
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description | This application claims priority under 35 U.S.C. §119 of German Patent Application, 10 2009 007 799.5, filed Feb. 6, 2009, which is incorporated herewith by reference in its entirety. The invention relates to a molecule for functionalizing the surface of an inert support and to the use in the preparation of a radionuclide of high purity in a generator. The invention relates in particular to a molecule for attaching a radioactive parent nuclide, in particular germanium-68, to a support. Radionuclides, in particular positron emitters, are used in positron emission tomography (PET). In the PET examination of a patient, the distribution of a weakly radioactive, positron emitter-labelled substance such as, for example, a biomolecule is visualized in an organism via the radioactive disintegration of the positron emitter, using a detector. Since biomolecules participate in the normal metabolism of the organism, accumulating in the process inter alia in tumour cells, PET can be utilized for identifying tumour cells. One example of a radionuclide preferred for PET is gallium-68, which can be obtained using a germanium-68/gallium-68 radionuclide generator system (1, 2). With a half-life of 67.63 minute's, the isotope gallium-68 disintegrates with emission of a positron. By virtue of its physical and chemical properties, gallium-68 is highly suitable for nuclear medical examinations. Owing to its short half-life, it is particularly suitable for radiolabelling biomolecules. Gallium-68 can be generated by radioactive disintegration from the parent nuclide germanium-68 which disintegrates with a half-life of 270.8 days. In the generator, the germanium-68 is attached to an insoluble matrix of an inert support, where, by continuous disintegration of the germanium, gallium-68 is constantly formed and may be extracted from the generator by elution with a solvent. The radionuclides used for labelling the radio-pharmaceutics have to meet high quality standards. In particular, the radionuclides generated have to have a high degree of purity and must be free from metallic impurities since these may, owing to competing reactions, have an adverse effect on the labelling of the radiopharmaceutics, and may reduce the technically achievable yield (3-5). As support for the stationary phase, known germanium-68/gallium-68 generator systems use inorganic ion exchange substances, such as, for example, TiO2, SnO2, Al(OH)3. However, in a disadvantageous manner, the gallium-68 extracted therewith contains metallic impurities, such that the original eluate has to be purified prior to use in a radiopharmaceutic (4, 5). As an alternative to inorganic ion exchange substances, generators use, as supports, organic polymers to which, with the aid of functional groups, individual molecules having a high affinity for germanium are attached. Such molecules may, for example, be pyrogallol or catechol which, via phenolic hydroxyl groups, form stable complexes with germanium (FIG. 1A) (6). In a known germanium-68/gallium-68 generator, the support used is a resin prepared from pyrogallol and formaldehyde (4-7). During the preparation of the germanium-specific resin, pyrogallol is immobilized on the support by copolymerization with formaldehyde. However, the applicability of these materials and generator systems is limited. Thus, with the germanium-68/gallium-68 generators mentioned above based on organic polymer, gallium-68 can be obtained only in concentrated acid solutions (3-6M). This requires reprocessing of the eluate prior to use as radiopharmaceutic. In addition, the process for synthesizing the pyrogallol/formaldehyde resin is technically very demanding and expensive. In addition, the main component of the formaldehyde matrix is toxic, such that the preparation of an injectable radiopharmaceutic requires additional purification steps. It was an object of the present invention to provide a substance for preparing a radionuclide using a generator, where a radionuclide can be attached to a support which can be used as stationary phase in the generator, and which allows the radionuclide to be prepared with a high degree of purity and without impurities, and also a corresponding generator and a preparation process. The object is achieved by a molecule having the features according to claim 1. According to the invention, a molecule for attaching a radioactive, parent nuclide to a support is provided which comprises at least one functional group for attaching the radioactive parent nuclide and a molecular moiety suitable for establishing nonpolar bonds to the support. By virtue of the nonpolar bond to the support, in an aqueous solution which can be used for eluting the daughter nuclide in a generator, the radioactive parent nuclide cannot be detached from the support material. It is thus possible to avoid contamination of the eluate, and to extend the lifespan of the generator for subsequent elutions. The radioactive parent nuclide may comprise germanium-which disintegrates to gallium-68. It is thus possible to provide a support for a germanium-68/gallium-68 generator which allows the preparation of highly pure gallium-68 substantially without impurities, in particular metallic impurities, and with a high degree of purity and preferably without further preparation steps prior to use in a radiopharmaceutic. The degree of purity that can be achieved is preferably less than 1 ppm, with preference less than 100 ppb, particularly preferably less than 10 ppb or even less than 1 ppb of impurities. According to one embodiment, the functional group for attaching the parent nuclide comprises a hydroxyl group and preferably a phenolic hydroxyl group. The molecule may also comprise a plurality of functional groups such as, for example, two, three or more functional groups. With the aid of the functional group, which has a high affinity to germanium, thus allowing quantitative adsorption of the germanium from the liquid phase, it is possible to form stable complexes with germanium molecules. According to a preferred embodiment, the parent nuclide is germanium-68 and the functional group is pyrogallol or catechol. According to a further preferred embodiment, the molecular moiety suitable for establishing a nonpolar bond to the support is hydrophobic. Using a hydrophobic molecular moiety, the molecule can be attached via a nonpolar bond to an inert support or be immobilized thereon, preventing inter alia a dissolution of the molecule and the parent nuclide attached thereto in an aqueous solution. In contrast, known compounds having one or more germanium-specific functional groups such as, for example, catechol and pyrogallol, are highly soluble in aqueous solutions. It is not possible to attach catechol and pyrogallol directly to an inert support such that the bond withstands extraction of the daughter nuclide from the generator using an aqueous solution. The solubility in water of catechol and pyrogallol is 450 g/l and 400 g/l, respectively. Using derivatives of molecules which, in addition to at least one germanium-specific functional group, additionally have a hydrophobic molecular moiety, it is possible to achieve insolubility in water. According to a further preferred embodiment, the hydro-phobic molecular moiety is selected from the group consisting of: (i) aromatic and heteroaromatic moieties, such as, for example, benzene, naphthalene, quinoline; (ii) saturated or unsaturated fatty acids having more than 3 carbon atoms, preferably from 3 to 20 carbon atoms; (iii) branched or straight-chain alkyl chains having more than 3 carbon atoms, such as, for example, octyl, decyl, or octadecyl groups, preferably having from 3 to 20 carbon atoms. According to a preferred embodiment, the molecule is an organic molecule selected from the group consisting of 2,3-dihydroxynaphthalene and dodecyl 3,4,5-trihydroxy-benzoate. According to a further preferred embodiment, the support is selected from the group consisting of an organic support and an inorganic support, such as, for example, silica gel. The invention furthermore provides a support for use as stationary phase, comprising at least one molecule according to the invention according to any of the embodiments described above and which is attached to the support via a nonpolar bond. The invention furthermore provides a generator for a radioactive daughter nuclide, in particular gallium-68, comprising a molecule according to the invention according to any of the embodiments described above, a support, the molecule being attached to the support via a nonpolar bond, and a parent nuclide, in particular germanium-68, which is attached to the molecule via the functional group. The invention furthermore provides a process for preparing a radioactive daughter nuclide which comprises the following steps: providing a generator comprising a support and a parent nuclide to which a molecule according to the invention according to any of the embodiments described above is attached, where the molecule is attached to the support via a nonpolar bond, and eluting the daughter nuclide. Using the generator according to the invention, it is possible to prepare gallium-68 with a particularly high degree of purity, metallic impurities and other residues from the generator being substantially avoidable. The generator can be prepared with low expense and in a cost-effective manner. According to a further preferred embodiment, the process comprises charging the generator with silica gel as support, to which silica gel the molecule is applied. According to yet another embodiment, the process comprises bringing the support into contact with the parent nuclide in a solution. Suitable solvents are the following substances: water, aqueous acids, solutions, salt solutions, such as, for example, buffer solutions, organic solutions based on alcohol, ether, etc. According to a further preferred embodiment, the parent nuclide may comprise germanium-68 which disintegrates to gallium-68. Finally, the invention comprises the use of a molecule according to any of the embodiments indicated above for preparing pure gallium-68. FIGS. 1a, b show the structural formulae of catechol (FIG. 1a) and pyrogallol as compounds having germanium-specific functional groups; FIGS. 2a, b show the structural formulae of examples of molecules according to the invention such as 2,3-dihydroxynaphthalene (FIG. 2a) and dodecyl 3,4,5-trihydroxybenzoate. A germanium-specific resin was prepared by coating inert silica gel with dodecyl 3,4,5-trihydroxybenzoate having the structural formula shown in FIG. 1b. The resin was used for preparing small chromatographic columns. An aqueous solution comprising the radio-nuclide germanium-68 having an activity in the range between 20 and 1250 MBq was then pumped through the columns. During this step, the germanium-68 was adsorbed quantitatively on the columns. The columns charged with germanium-68 were then used to prepare the short-lived gallium-68. It was possible to repeatedly eluate the gallium-68 generated by the germanium-68 absorbed on the support. The elution of gallium-68 was effected using weak hydrochloric acid solutions (0.05 M HCl) having a low volume of up to 2.5 ml. Leakage of the parent nuclide germanium-68 was in the range 1×10−4−3×10−3%. The gallium-68 could be used directly and without further chemical reprocessing for preparing injectable gallium-68 radio-pharmaceutics. Literature References: 1) Al-Nahhas A, Win Z, Szysko T, Singha A, Nannil C, Fanti S, Rubello D. Gallium-68 PET: A New Frontier in Receptor Cancer Imaging. Anticancer research. 2007; 27: 4087-4094 2) Helmut M, Hofmann M, Haberkorn U. 68Ga-Labeled Peptides in Tumor Imaging. J Nuc Med. 2005; 46: 172S-178S 3) Breeman W, Jong M, Blois E, Bernard B, Konijnenberg M, Krenning E. Radiolabelling DOTA-peptides with 68Ga. Eur J Nuc Med Mol Imaging. 2005; 32: 478-458 4) Meyer G-J, Mäcke H, Schuhmacher J, Knapp W, Hofmann M. 68Ga-labelled DOTA-derivatised peptide ligands. Eur J Nuc Med Mol Imaging. 2004; 31: 1097-1104 5) Zhernosekov K, Filosofov D, Baum R, Aschoff P, Bihl H, Razbash A, Jahn M, Jennewein M, Rösch F. Processing of generator-produced 68Ga for medical application. J Nuc Med. 2007; 48: 1741-1748 6) Patent DE 29 32 948 A1 7) Schuhmacher J, Maier-Borst W. A new 68Ge/68Ga radio-isotope generator system for production of 68Ga in dilute HCl. I J Appl Rad Isotopes. 1981; 32: 31-36 |
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054835607 | summary | BACKGROUND OF THE INVENTION FIELD OF THE INVENTION The invention relates to a process for testing, repairing or exchanging nozzles penetrating the bottom of a reactor pressure vessel, wherein each nozzle serves for the introduction of a respective probe of an in-core instrumentation of a water-cooled nuclear reactor. Depending on the power and type of the nuclear reactor, a greater or lesser number (about 50) of nozzles pass through the bottom of the reactor pressure vessel. The probes, which are fitted with detectors on their free end regions, are made to enter the fuel assemblies during reactor operation. They are therefore exposed to a high level of radioactive radiation. If testing, repairing or exchanging of nozzles is necessary, the fuel assemblies of the reactor core and all of the core internals are removed from the reactor pressure vessel. Before taking out the fuel assemblies, the probes of the in-core instrumentation, which can travel in channels of the fuel assemblies, must be parked outside in guide tubes leading away from the nozzles. Since the detectors are highly contaminated, due to their insertion in the reactor core, an undesired radioactive loading is the consequence. SUMMARY OF THE INVENTION It is accordingly an object of the invention to provide a method and an apparatus for testing, repairing or exchanging nozzles passing through the bottom of a reactor pressure vessel, which overcome the hereinafore-mentioned disadvantages of the heretofore-known methods and devices of this general type and in which radioactive loading by detectors of in-core instrumentation is considerably reduced. With the foregoing and other objects in view there is provided, in accordance with the invention, in a method for testing, repairing or exchanging nozzles penetrating a bottom of a reactor pressure vessel, wherein each nozzle serves for the introduction of a respective probe of an in-core instrumentation of a water-cooled nuclear reactor, the improvement which comprises: a. inserting at least one shielding container into the reactor pressure vessel; PA1 b. introducing all of the probes of the in-core instrumentation into a shielding region of the at least one shielding container; PA1 c. withdrawing at least one of the probes assigned to one of the nozzles to be tested or worked on, from the reactor pressure vessel; PA1 d. exposing a shaft extending through the shielding container and being assigned to at least one of the nozzles; PA1 e. introducing testing devices or tools through the shaft; PA1 f. carrying out at least one of testing and working operations; PA1 g. withdrawing the testing devices or tools from the shaft; PA1 h. shielding the shaft; PA1 i. introducing the probe into the shielding region of the shielding container; and PA1 j. repeating steps c to j for further nozzles. During work being performed on a nozzle assigned to one shaft, the remaining probes are located in the shielding region of the shielding container, so that the radioactive loading inside and outside the pressure vessel is reduced considerably. The method successfully accomplishes an integration of the shielding and working operations. In accordance with another mode of the invention, there is provided a method which comprises cutting off a nozzle to be repaired, just above an inner surface of the bottom; drilling open a nozzle part remaining in the bottom to form a staged nozzle bore; inserting a new staged nozzle part into the nozzle bore; establishing a supporting connection by means of a weld seam; and applying a sealing weld seam at a lower end of the staged bore. With the objects of the invention in view, there is also provided, in a water-cooled nuclear reactor having a reactor pressure vessel with a bottom, a flange and a cross section, and nozzles each penetrating the bottom for the introduction of a respective probe of an in-core instrumentation of the nuclear reactor, an apparatus for testing, repairing or exchanging the nozzles, comprising a plurality of shielding containers together filling the cross section of the reactor pressure vessel, the shielding containers having a supporting flange supported on the reactor pressure vessel flange, and the shielding containers having bottom plates ending above the nozzles; a plurality of shafts passing through the bottom plate of at least one of the shielding container and engaging over the nozzles located in a corresponding partial cross section of the reactor pressure vessel; a water-filled cartridge to be inserted into each of the shafts, the cartridges having bottoms each ending above a respective one of the nozzles; tubes for receiving the probes; at least one of the tubes being embedded in at least one of the bottoms and having a closed end protruding by a predetermined amount into the cartridge and an open end being adjacent one of the nozzles; the bottom plate of the others or the remainder of the shielding containers through which the shafts do not pass, having others of the tubes embedded therein with closed ends protruding by a predetermined amount into the shielding containers and open ends adjacent the nozzles covered by or facing away from the shielding container; carriers for traveling in the shaft and receiving at least one of testing devices and tools; and each of the shafts having means for guiding one of the carriers. The inside diameters of the tubes correspond approximately to that of the nozzles, so that they serve as guide tubes for the probes and so that their free ends, which are fitted with the detectors, pass without any problem into the shielding region of the shielding container. As a rule, the cross section of the reactor pressure vessel is covered by four shielding containers, each including approximately a quadrant sector. If only one quadrant sector is provided with shafts, after the completion of work in one reactor pressure vessel sector, the shielding container can be turned through 90.degree. in order to work on the nozzles of the next reactor pressure vessel sector. In the shafts which are not accessible at that time to the carrier for receiving testing devices or tools, a water-filled cartridge performs the shielding task. Consequently, an apparatus has been created in which the working and shielding are combined in a compact structure. In accordance with another feature of the invention, the means for guiding the carrier are preferably assigned to the region of the shaft beginning underneath the respective cartridge. Consequently, the guiding means do not result in any cross-sectional reduction of the cartridges and consequently do not result in any reduction in the shielding effect. In accordance with a further feature of the invention, the guiding means are rails on which the carrier can be securely clamped. In accordance with a concomitant feature of the invention, there is provided a sealing element disposed between the bottom of the reactor pressure vessel and the lower edge of the shaft, for localizing the swarf or chips generated inside a shaft. Other features which are considered as characteristic for the invention are set forth in the appended claims. Although the invention is illustrated and described herein as embodied in a method and an apparatus for testing, repairing or exchanging nozzles passing through the bottom of a reactor pressure vessel, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims. The construction and method of operation of the invention, however, 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. |
063103532 | summary | TECHNICAL FIELD OF THE INVENTION The present invention relates generally to the mass spectroscopic analysis of chemical samples and more particularly to time-of-flight mass spectrometry. More specifically, a means and method are described for the focusing of ions in time-of-flight mass spectrometry. BACKGROUND OF THE INVENTION This invention relates in general to ion beam handling in mass spectrometers and more particularly to a means of focusing ions in time-of-flight mass spectrometers (TOFMS). The apparatus and method of mass analysis described herein is an enhancement of the techniques that are referred to in the literature relating to mass spectrometry. The analysis of ions by mass spectrometers is important, as mass spectrometers are instruments that are used to determine the chemical structures of molecules. In these instruments, molecules become positively or negatively charged in an ionization source and the masses of the resultant ions are determined in vacuum by a mass analyzer that measures their mass/charge (m/z) ratio. Mass analyzers come in a variety of types, including magnetic field (B), combined (double-focusing) electrical (E) and magnetic field (B), quadrupole (Q), ion cyclotron resonance (ICR), quadrupole ion storage trap, and time-of-flight (TOF) mass analyzers, which are of particular importance with respect to the invention disclosed herein. Each mass spectrometric method has a unique set of attributes. Thus, TOFMS is one mass spectrometric method that arose out of the evolution of the larger field of mass spectrometry. The analysis of ions by TOFMS is, as the name suggests, based on the measurement of the flight times of ions from an initial position to a final position. Ions which have the same initial kinetic energy but different masses will separate when allowed to drift through a field free region. Ions are conventionally extracted from an ion source in small packets. The ions acquire different velocities according to the mass-to-charge ratio of the ions. Lighter ions will arrive at a detector prior to high mass ions. Determining the time-of-flight of the ions across a propagation path permits the determination of the masses of different ions. The propagation path may be circular or helical, as in cyclotron resonance spectrometry, but typically linear propagation paths are used for TOFMS applications. TOFMS is used to form a mass spectrum for ions contained in a sample of interest. Conventionally, the sample is divided into packets of ions that are launched along the propagation path using a pulse-and-wait approach. In releasing packets, one concern is that the lighter and faster ions of a trailing packet will pass the heavier and slower ions of a preceding packet. Using the traditional pulse-and-wait approach, the release of an ion packet as timed to ensure that the ions of a preceding packet reach the detector before any overlap can occur. Thus, the periods between packets is relatively long. If ions are being generated continuously, only a small percentage of the ions undergo detection. A significant amount of sample material is thereby wasted. The loss in efficiency and sensitivity can be reduced by storing ions that are generated between the launching of individual packets, but the storage approach carries some disadvantages. Resolution is an important consideration in the design and operation of a mass spectrometer for ion analysis. The traditional pulse-and-wait approach in releasing packets of ions enables resolution of ions of different masses by separating the ions into discernible groups. However, other factors are also involved in determining the resolution of a mass spectrometry system. "Space resolution" is the ability of the system to resolve ions of different masses despite an initial spatial position distribution within an ion source from which the packets are extracted. Differences in starting position will affect the time required for traversing a propagation path. "Energy resolution" is the ability of the system to resolve ions of different mass despite an initial velocity distribution. Different starting velocities will affect the time required for traversing the propagation path. In addition, two or more mass analyzers may be combined in a single instrument to form a tandem mass spectrometer (MS/MS, MS/MS/MS, etc.). The most common MS/MS instruments are four sector instruments (EBEB or BEEB), triple quadrupoles (QQQ), and hybrid instruments (EBQQ or BEQQ). The mass/charge ratio measured for a molecular ion is used to determine the molecular weight of a compound. In addition, molecular ions may dissociate at specific chemical bonds to form fragment ions. Mass/charge ratios of these fragment ions are used to elucidate the chemical structure of the molecule. Tandem mass spectrometers have a particular advantage for structural analysis in that the first mass analyzer (MS1) can be used to measure and select molecular ion from a mixture of molecules, while the second mass analyzer (MS2) can be used to record the structural fragments. In tandem instruments, a means is provided to induce fragmentation in the region between the two mass analyzers. The most common method employs a collision chamber filled with an inert gas, and is known as collision induced dissociation CID. Such collisions can be carried out at high (5-10 keV) or low (10-100 eV) kinetic energies, or may involve specific chemical (ion-molecule) reactions. Fragmentation may also be induced using laser beams (photodissociation), electron beams (electron induced dissociation), or through collisions with surfaces (surface induced dissociation). It is possible to perform such an analysis using a variety of types of mass analyzers including TOF mass analysis. In a TOFMS instrument, molecular and fragment ions formed in the source are accelerated to a kinetic energy EQU eV=1/2mv.sup.2 (1) where e is the elemental charge, V is the potential across the source/accelerating region, m is the ion mass, and v is the ion velocity. These ions pass through a field-free drift region of length L with velocities given by equation 1. The time required for a particular ion to traverse the drift region is directly proportional to the square root of the mass/charge: ##EQU1## Conversely, the mass/charge ratios of ions can be determined from their flight times according to the equation: ##EQU2## where a and b are constants which can be determined experimentally from the flight times of two or more ions of known mass/charge ratios. Generally, TOF mass spectrometers have limited mass resolution. This arises because there may be uncertainties in the time that the ions were formed (time distribution) a in their location in the accelerating field at the time they were formed (spatial distribution), and in their initial kinetic energy distributions prior to acceleration (energy distribution). The first commercially successful TOFMS was based on an instrument described by Wiley and McLaren in 1955 (Wiley, W. C.; McLaren, I. H., Rev. Sci. Instrumen. 26 1150 (1955)). That instrument utilized electron impact (EI) ionization (which is limited to volatile samples) and a method for spatial and energy focusing known as time-lag focusing. In brief, molecules are first ionized by a pulsed (1-5 microsecond) electron beam. Spatial focusing was accomplished using multiple-stage acceleration of the ions. In the first stage, a low voltage (-150 V) draw-out pulse is applied to the source region that compensates for ions formed at different locations, while the second (and other) stages complete the acceleration of the ions to their final kinetic energy (-3 keV ).A short time-delay (1-7 microseconds) between the ionization and draw-out pulses compensates for different initial kinetic energies of the ions, and is designed to improve mass resolution. Because this method required a very fast (40 ns) rise time pulse in the source region, it was convenient to place the ion source at ground potential, while the drift region floats at -3 kV. The instrument was commercialized by Bendix Corporation as the model NA-2, and later by CVC Products (Rochester, N.Y.) as the model CVC-2000 mass spectrometer. The instrument has a practical mass range of 400 Daltons and a mass resolution of 1/300, and is still commercially available. There have been a number of variations on this instrument. Muga (TOFTEC, Gainsville) has described a velocity compaction technique for improving the mass resolution (Muga velocity compaction). Chatfield et al. (Chatfield FT-TOF) described a method for frequency modulation of gates placed at either end of the flight tube, and Fourier transformation to the time domain to obtain mass spectra. This method was designed to improve the duty cycle. Cotter et al. (VanBreeman, R. B.: Snow, M.: Cotter, R. J., Int. J. Mass Spectrom. Ion Phys. 49 (1983) 35.; Tabet, J. C.; Cotter, R. J., Anal. Chem. 56 (1984) 1662; Olthoff, J. K.; Lys, I.: Demirev, P.: Cotter, R. J., Anal. Instrumen. 16 (1987) 93, modified a CVC 2000 time-of-flight mass spectrometer for infrared laser desorption of involatile biomolecules, using a Tachisto (Needham, Mass.) model 215G pulsed carbon dioxide laser. This group also constructed a pulsed liquid secondary time-of-flight mass spectrometer (liquid SIMS-TOF) utilizing a pulsed (1-5 microsecond) beam of 5 keV cesium ions, a liquid sample matrix, a symmetric push/pull arrangement for pulsed ion extraction (Olthoff, J. K.; Cotter, R. J., Anal. Chem. 59 (1987) 999-1002.; Olthoff, J. K.; Cotter, R. J., Nucl. Instrum. Meth. Phys. Res. B-26 (1987) 566-570. In both of these instruments, the time delay range between ion formation and extraction was extended to 5-50 microseconds, and was used to permit metastable fragmentation of large molecules prior to extraction from the source. This in turn reveals more structural information in the mass spectra. The plasma desorption technique introduced by Macfarlane and Torgerson in 1974 (Macfarlane, R. D.; Skowronski, R. P.; Torgerson, D. F., Biochem. Biophys. Res Commoun. 60 (1974) 616.) formed ions on a planar surface placed at a voltage of 20 kV. Since there are no spatial uncertainties, ions are accelerated promptly to their final kinetic energies toward a parallel, grounded extraction grid, and then travel through a grounded drift region. High voltages are used, since mass resolution is proportional to U o/;eV, where the initial kinetic energy, U o/is of the order of a few electron volts. Plasma desorption mass spectrometers have been constructed at Rockefeller (Chait, B. T.; Field, F. H., J. Amer. Chem. Soc. 106 (1984) 193), Orsay (LeBeyec, Y.; Della Negra, S.; Deprun, C.; Vigny, P.; Giont, Y. M., Rev. Phys. Appl 15 (1980) 1631), Paris (Viari, A.; Ballini, J. P.; Vigny, P.; Shire, D.; Dousset, P., Biomed. Environ. Mass Spectrom, 14 (1987) 83), Upsalla (Hakansson, P.; Sundqvist B., Radiat. Eff. 61 (1982) 179) and Darmstadt (Becker, O.; Furstenau, N.; Krueger, F. R.; Weiss, G.; Wein, K., Nucl. Instrum. Methods 139 (1976) 195). A plasma desorption time-of-flight mass spectrometer has been commercialized by BIO-ION Nordic (Upsalla, Sweden). Plasma desorption utilizes primary ion particles with kinetic energies in the MeV range to induce desorption/ionization. A similar instrument was constructed at Manitobe (Chain, B. T.; Standing, K. G., Int. J. Mass Spectrum. Ion Phys. 40 (1981) 185) using primary ions in the keV range, but has not been commercialized. Matrix-assisted laser desorption, introduced by Tanaka et al. (Tanaka, K.; Waki, H.; Ido, Y.; Akita, S.; Yoshida, Y.; Yoshica, T., Rapid Commun. Mass Spectrom. 2 (1988) 151) and by Karas and Hillenkamp (Karas, M.; Hillenkamp, F., Anal. Chem. 60 (1988) 2299) utilizes TOFMS to measure the molecular weights of proteins in excess of 100,000 Daltons. An instrument constructed at Rockefeller (Beavis, R. C.; Chait, B. T., Rapid Commun. Mass Spectrom. 3 (1989) 233) has been commercialized by VESTEC (Houston, Tex), and employs prompt two-stage extraction of ions to an energy of 30 keV. Time-of-flight instruments with a constant extraction field have also been utilized with multi-photon ionization, using short pulse lasers. The instruments described thus far are linear time-of-flights, that is--there is no additional focusing after the ions are accelerated and allowed to enter the drift region. Two approaches to additional energy focusing have been utilized--those which pass the ion beam through an electrostatic energy filter. The reflectron (or ion mirror) was first described by Mamyrin (Mamyrin, B. A.; Karatajev, V. J.; Shmikk, D. V.; Zagulin, V. A., Sov. Phys., JETP 37 (1973) 45). At the end of the drift region, ions enter a retarding field from which they are reflected back through the drift region at a slight angle. Improved mass resolution results from the fact that ions with larger kinetic energies must penetrate the reflecting field more deeply before being turned around. These faster ions than catch up with the slower ions at the detector and are focused. Reflectrons were used on the laser microprobe instrument introduced by Hillenkamp et al. (Hillenkamp, F.; Kaufmann, R.; Nitsche, R.; Unsold, E., Appl. Phys. 8 (1975) 341) and commercialized by Leybold Hereaus as the LAMMA (LAser Microprobe Mass Analyzer). A similar instrument was also commercialized by Cambridge Instruments as the IA (Laser Ionization Mass Analyzer). Benninghoven (Benninghoven reflectron) has described a SIMS (secondary ion mass spectrometer) instrument that also utilizes a reflectron, and is currently being commercialized by Leybold Hereaus. A reflecting SIMS instrument has also been constructed by Standing (Standing, K. G.; Beavis, R.; Bollbach, G.; Ens, W.; Lafortune, F.; Main, D.; Schueler, B.; Tang, X.; Westmore, J. B., Anal. Instrumen. 16 (1987) 173). Lebeyec (Della-Negra, S.; Lebeyec, Y., in Ion Formation from Organic Solids IFOS III, ed. by A. Benninghoven, pp 42-45, Springer-Verlag, Berlin (1986)) described a coaxial reflectron time-of-flight that reflects ions along the same path in the drift tube as the incoming ions, and records their arrival times on a channelplate detector with a centered hole that allows passage of the initial (unreflected) beam. This geometry was also utilized by Tanaka et al. (Tanaka, K.; Waki, H.; Ido, Y.; Akita, S.; Yoshida, T., Rapid Comun. Mass Spectrom. 2 (1988) 151) for matrix assisted laser desorption. Schlag et al. (Grotemeyer, J.; Schlag, E. W., Org. Mass Spectrom. 22 (1987) 758) have used a reflectron on a two-laser instrument. The first laser is used to ablate solid samples, while the second laser forms ions by multi-photon ionization. This instrument is currently available from Bruker. Wollnik et al. (Grix., R.; Kutscher, R.; Li, G.; Gruner, U.; Wollnik, H., Rapid Commun. Mass Spectrom. 2 (1988) 83) have described the use of reflectrons in combination with pulsed ion extraction, and achieved mass resolutions as high as 20,000 for small ions produced by electron impact ionization. An alternative to reflectrons is the passage of ions through an electrostatic energy filter, similar to that used in double-focusing sector instruments. This approach was first described by Poschenroeder (Poschenroeder, W., Int. J. Mass Spectrom. Ion Phys. 6 (1971) 413). Sakurai et al. (Sakuri, T.; Fujita, Y; Matsuo, T.; Matsuda, H; Katakuse, I., Int. J. Mass Spectrom. Ion Processes 66 (1985) 283) have developed a time-of-flight instrument employing four electrostatic energy analyzers (ESA) in the time-of-flight path. At Michigan State, an instrument known as the ETOF was described that utilizes a standard ESA in the TOE analyzer (Michigan ETOF). Lebeyec et al. (Della-Negra, S.; Lebeyec, Y., in Ion Formation from Organic Solids IFOS III, ed. by A. Benninghoven, pp 42-45, Springer-Verlag, Berlin (1986)) have described a technique known as correlated reflex spectra, which can provide information on the fragment ion arising from a selected molecular ion. In this technique, the neutral species arising from fragmentation in the flight tube are recorded by a detector behind the reflectron at the same flight time as their parent masses. Reflected ions are registered only when a neutral species is recorded within a preselected time window. Thus, the resultant spectra provide fragment ion (structural) information for a particular molecular ion. This technique has also been utilized by Standing (Standing, K. G.; Beavis, R.; Bollbach, G.; Ens, W.; Lafortune, F.; Main, D.; Schueler, B.; Tang, X.; Westmore, J. B., Anal. Instrumen. 16 (1987) 173). Although TOF mass spectrometers do not scan the mass range, but record ions of all masses following each ionization event, this mode of operation has some analogy with the linked scans obtained on double-focusing sector instruments. In both instruments, MS/MS information is obtained at the expense of high resolution. In addition, correlated reflex spectra can be obtained only on instruments which record single ions on each TOF cycle, and are therefore not compatible with methods (such as laser desorption) which produce high ion currents following each laser pulse. New ionization techniques, such as plasma desorption (Macfarlane, R. D.; Skowronski, R. P.; Torgerson, D. F.; Biochem. Bios. Res. Commun. 60 (1974) 616), laser desorption (VanBreemen, R. B.; Snow, M.; Cotter, R. J., Int. J. Mass Spectrom. Ion Phys. 49 (1983) 35; Van der Peyl, G. J. Q.; Isa, K.; Haverkamp, J.; Kistemaker, P.G., Org. Mass Spectrom. 16 (1981) 416), fast atom bombardment (Barber, M.; Bordoli, R. S.; Sedwick, R. D.; Tyler, A. N., J. Chem. Soc., Chem. Commun. (1981) 325-326) and electrospray (Meng, C. K.; Mann, M.; Fenn, J. B., Z. Phys. D10 (1988) 361), have made it possible to examine the chemical structures of proteins and peptides, glycopeptides, glycolipids and other biological compounds without chemical derivatization. The molecular weights of intact proteins can be determined using matrix assisted laser desorption ionization (MALDI) on a TOF mass spectrometer or electrospray ionization. For more detailed structural analysis, proteins are generally cleaved chemically using CNBr or enzymatically using trypsin or other proteases. The resultant fragments, depending upon size, can be mapped using MALDI, plasma desorption or fast atom bombardment. In this case, the mixture of peptide fragments (digest) is examined directly resulting in a mass spectrum with a collection of molecular ion corresponding to the masses of each of the peptides. Finally, the amino acid sequences of the individual peptides which make up the whole protein cain be determined by fractionation of the digest, followed by mass spectral analysis of each peptide to observe fragment ions that correspond to its sequence. It is the sequencing of peptides for which tandem mass spectrometry has its major advantages. Generally, most of the new ionization techniques are successful in producing intact molecular ions, but not in producing fragmentation. In a tandem instrument the first mass analyzer passes molecular ions corresponding to the peptide of interest. These ions are activated toward fragmentation in a collision chamber, and their fragmentation products are extracted and focused into the second mass analyzer which records a fragment ion (or daughter ion) spectrum. A tandem TOFMS consists of two TOF analysis regions with an ion gate between the two regions. The ion gate allows one to gate (i.e. select) ions which will be passed from the first TOF analysis region to the second. As in conventional TOFMS, ions of increasing mass have decreasing velocities and increasing flight times. Thus, the arrival time of ions at the ion gate at the end of the first TOF analysis region is dependent on the mass-to-charge ratio of the ions. If one opens the ion gate only at the arrival time of the ion mass of interest, then only ions of that mass-to-charge will be passed into the second TOF analysis region. However, it should be noted that the products of an ion dissociation that occurs after the acceleration of the ion to its final potential will have the same velocity as the original ion. The product ions will therefore arrive at the ion gate at the same time as the original ion and will be passed by the gate (or not) just as the original ion would have been. The arrival times of product ions at the end of the second TOF analysis region is dependent on the product ion mass because a reflectron is used. As stated above, product ions have the same velocity as the reactant ions from which they originate. As a result, the kinetic energy of a product ion is directly proportional to the product ion mass. Because the flight time of an ion through a reflectron is dependent on the kinetic energy of the ion, and the kinetic energy of the product ions are dependent on their masses, the flight time of the product ions through the reflectron is dependent on their masses. In TOF mass spectrometers, it is often the case that divergent ion beams are produced by the ion source. As a result, the ion beam must be focused by an electrostatic lens in order for a large fraction (i.e. >50%) of the ions to be successfully analyzed and detected. However, prior art lenses tend to operate at high voltage--often in excess of 10 kV--and have a significant influence on the flight time of the ions being analyzed. Further, the lens may have an unwanted influence on the ion's peak shape or width. The purpose of the present invention is to focus ions onto a detector or other device with improved efficiency and decreased influence on the ion's flight time at a lower operating voltage than prior art devices. Several references relate to the technology herein disclosed. For example, F. Hillenkamp, M. Karas, R. C. Beavis, B. T. Chait, Anal. Chem. 63(24), 1193A(1991); Wei Hang, Pengyuan Yag, Xiaoru Wang, Chenglong Yang, Yongxuan Su, and Benli Huang, Rapid Comm. Mass Spectrom. 8, 590(1994); A. N. Verentchikov, W. Ens, K. G. Standing, Anal.Chem. 66, 126(1994); J. H. J. Dawson, M. Guilhaus, Rapid Comm. Mass Spectrom. 3, 155(1989); M. Guilhaus, J. Am. Soc. Mass Spectrom. 5, 588(1994); E. Axelsson, L. Holmlid, Int. J. Mass Spectrom. Ion Process. 59, 231(1984); O. A. Mirgorodskaya, et al., Anal. Chem. 66, 99(1994); S. M. Michael, B. M. Chien, D. M. Lubman, Anal. Chem. 65, 2614(1993); W. C. Wiley, I. H. McLaren, Rev. Sci. Inst. 26(12), 1150(1955). SUMMARY OF THE INVENTION The present invention relates in general to ion beam handling in mass spectrometers and more particularly to a means of focusing ions in time-of-flight mass spectrometers (TOFMS). Electrostatic ion lenses rely on the deflection of ions by electrostatic fields to accomplish focusing. Among other factors, the intensity of the electrostatic field produced by a lens as a function of position determines the degree of deflection which ions passing through the device will experience. The direction of motion of ions entering an ion lens is typically related to the position at which the ion enters the lens. Proper focusing is obtained by forming an electrostatic field which deflects ions by varying degrees according to their entrance position in the same manner that the ion's direction of motion varies with entrance position. A properly formed electrostatic field can be used to decrease the angle of divergence of an ion beam or focus it to a focal point. However, the electrostatic field of a lens may also produce unwanted effects. This electrostatic field will accelerate ions in the primary direction of ion motion as well as perpendicular to it. As a result, the time that ions spend in the lens will vary depending on their entrance position. In TOFMS, the result is a variation of the flight time of the ions and a broadening of the detection signal. In the prior art, unshielded lenses are used wherein the electrostatic field is formed in such a way that the ion beam is actually defocused by portions of the field and focused by others. This inefficient use of the electrostatic field results in a higher operating voltage and more intense undesired effects than would otherwise be necessary. A lens according to the present invention is shielded so as to eliminate the defocusing portions of the electrostatic field. This reduces the operating voltage by as much as an order of magnitude and thereby reduces many of the unwanted effects observed with conventional electrostatic lenses. The present invention is a specific design for a TOF mass spectrometer incorporating laser desorption, and a patented (United States Patent No. 4,731,532) two stage gridless reflector. Other objects, features, and characteristics of the present invention, as well as the methods of operation and functions of the related elements of the structure, and the combination of parts and economies of manufacture, will become more apparent upon consideration of the following detailed description with reference to the accompanying drawings, all of which form a part of this specification. |
051006103 | abstract | System for inspecting the interior of tube plugs, particularly tube plugs of the type having expander elements disposed therein. The system includes a probe assembly connectable to the tube plug, the probe assembly having a rotatable and translatable probe carrier housing for housing a sensor probe capable of inspecting the tube plug. The probe assembly also includes a leaf spring extending through the probe carrier housing and attached to the sensor probe for radially outwardly biasing the sensor probe into sensing contact with the interior of the tube plug, particularly the interior of the tube plug located between the top of the tube plug and the expander element. A tangle-free flexible hose is connected at one end thereof to the probe carrier housing for transversely rotating and longitudinally translating the probe carrier housing. The flexible hose is also connected at the other end thereof to a probe driver assembly that rotates the flexible hose which in turn rotates the probe carrier housing. Rotation of the probe carrier housing rotatably translates the probe carrier housing into the tube plug for inspecting the tube plug in a helical scanning pattern. |
summary | ||
050911465 | summary | This invention relates to nuclear fuel bundles having part length rods. More particularly, part length rods are disclosed in combination with overlying steam vent tubes for permitting the high velocity localized discharge of steam from the fuel bundle. A system of apertures and devices is disclosed distributed along the steam vent tube to preferentially enable steam flow interior of the steam vent tube and preferentially exclude liquid moderator flow interior of the steam vent tube. BACKGROUND OF THE INVENTION Ueda Japanese Patent Showa 52-50489 disclosed the use of part length fuel rods for the creation of an improved fuel to moderator ratio in the upper two phase region of a fuel bundle, especially in the cold shut down state of the reactor. A fuel bundle was disclosed in which clustered part length fuel rods defined a large, central, generally conical shaped void in the upper two phase region of the fuel bundle. Two embodiments were disclosed in the Ueda reference. A first embodiment includes a large conical water rod occupying the large central conical void defined collectively by the part length rods. A second, and apparently preferred embodiment, disclosed the conical region otherwise unoccupied. Regarding this latter design, testing has established that while nuclear improvements in the upper two phase region in the cold state might be realized, adverse heat transfer performance, especially in terms of adverse critical power may be realized by large central void regions in a boiling water reactor fuel bundle. Specifically, the large defined void results in vapor being concentrated to the region. Unfortunately, surrounding portions of the two phase region tends to flow into the steam vent area. This results in the diversion of significant amounts of liquid coolant away from the heated rod surfaces adjacent the void, this liquid coolant being entrained in the accelerated steam flow within the large defined conical void. There results a reduced flow adjacent the full length fuel rods which surround the large void. This reduced flow rate has a corresponding reduced critical power on the rod surfaces adjacent the void. Overall fuel bundle efficiency is reduced. In Dix et al. U.S. Pat. No. 5,017,332 entitled Two-Phase Pressure Drop Reduction BWR Assembly Design, issued May 21, 1991 (formerly U.S. Pat. Application Ser. No. 07/176,975, filed Apr. 4, 1988), we maintained the nuclear benefits and removed the adverse thermal hydraulic effects by using generally smaller open flow channels dispersed across the two phase region of the fuel bundle lattice. The dispersed flow channels realized the natural tendency of the vapor phase of the two phase mixture to migrate ("drift") toward the low resistance flow paths where the realized flow was primarily vapor. It has been found that such dispersed flow paths are favorable for a BWR fuel assembly since preferential diversion of vapor away from the fueled rods has combined nuclear, stability, and thermal hydraulic advantages. SUMMARY OF THE INVENTION In a fuel bundle for a boiling water reactor having one or more part length rods in the two phase region, one or more steam vent tubes are introduced overlying the part length rods. The fuel bundle includes a lower tie plate for admitting water moderator and supporting a plurality of fuel rods in upstanding side-by-side relation, an upper tie plate for permitting water and steam to be discharged from the top of the fuel bundle and maintaining the fuel rods in upstanding side-by-side relation, a surrounding fuel channel for confining moderator flow along a path over the fuel rods and between the tie plates, and dispersed vertically intermittent spacers for maintaining the fuel rods in their designed side by side relation. One or more fuel rods extends from the lower tie plate vertically less than the full length to the upper tie plate ending interior of the fuel bundle at a disposition where the upper end of the part length rods is braced in the vertical position by a spacer. At least one of these partial length rods is provided with an overlying steam vent tube. This steam vent tube has openings and devices distributed along its length to encourage steam flow interior of the tube and remove liquid flow from the interior of the tube. The vent tube wall eliminates the interface drag between the steam interior of the tube and the surrounding water steam mixture. The presence of the steam vent tube as a high velocity steam escape path enables remaining portions of the fuel bundle to contain a higher liquid moderator fraction with flatter axial voids and power distributions during the operating state of the fuel bundle. The presence of high velocity escaping steam combined with distributed apertures along the length of the steam vent tube promotes resistance to fluid oscillations within the fuel bundles. Variations of steam vent tubes are disclosed including steam vent tubes overlying multiple clustered part length rods. There results a fuel bundle design in which the entire upper cross section of the fuel bundle is devoted to steam generation and coolant outflow as contrasted to a fuel bundle having conventional water rods. OTHER OBJECTS, FEATURES AND ADVANTAGES An object of this invention is to disclose the use of at least one steam vent tube overlying at least one part length fuel rod within a fuel bundle of a boiling water nuclear reactor. According to this aspect, the part length rod--resting on the lower tie plate and terminating before reaching the upper tie plate--is provided with an overlying tube. This tube opens overlying the end of the part length rod and extends from the vicinity of the end of the part length fuel rod at least part way the distance between the end of the fuel rod and the upper tie plate. Typically, the tube is open overlying the part length rods and open at its point of discharge to and through the upper tie plate. Because of the opening of the tube overlying the end of the part length rod, the steam vent tube preferentially admits initial vapor flow to the interior of the tube and thereafter rapidly vents the vapor to discharge at the upper tie plate. An advantage of the steam vent tube is that it forms a dividing wall flow interface between rapidly flowing, upwardly venting steam interior of the tube and the surrounding two phase steam water mixture exterior of the steam vent tube. There results an isolated steam flow channel having a reduction in flow friction and a local increase in the volumetric flux of vapor outflow. The turbulent interface between the upwardly flowing steam and the surrounding two phase mixture is removed. Consequently, entrainment between the upwardly flow vapor and the bounding two phase liquid vapor region is removed. A further advantage of the steam vent tube is that it establishes a localized high velocity steam vent within the interior upper two phase region of the fuel bundle. As a result, less of the upper cross sectional area of the fuel bundle within the two phase region is required for upward vapor flow. The upper volume of the fuel bundle not occupied by vapor flow can instead be occupied by two phase liquid vapor mixture having a higher water content. There results a fuel bundle capable of maintaining a higher overall liquid moderator content. Flatter axial voids and power distributions result. Yet an additional advantage of the steam vent tube is that the entire upper portion of the fuel bundle is utilized either for the generation of steam or out flow of liquid and vapor moderator. Large flow blocking water rods for increasing the fuel to moderator ratio--but also otherwise obstructing flow area--are not required. A fuel bundle concept is disclosed in which the steam vent tube in classifying water to the exterior of the vent tube functions to both provide improved moderator fraction while maintaining (and not obstructing) flow area. An additional advantage of the disclosed concept is that it provides additional moderator in a manner which is the inverse of moderator provided by the more typically used large water rods. By the expedient of providing a concentrated isolated flow path for the upward preferential flow of vapor, the remain portions of the fuel bundle in the upper two phase region can have increase liquid moderator content--equivalent in may cases to that moderator content supplied by a large water rod. At the same time, all areas of the upper two phase region of the bundle are available for moderator flow. As a result, no portion in the fuel bundle need be dedicated to flow obstructing water rods. An additional object of this invention is to disclose a series of openings along the length of the steam vent tubes, these opening designed to preferentially admit vapor interior of the tubes and to preferentially exclude liquid interior of the tube. According to this aspect of the invention, the steam vent tube has at least one--and is preferably manifolded--with openings along its length. These openings are designed with vapor liquid separation configurations for preferentially admitting vapor interior of the steam vent tube and preferentially excluding liquid interior of the steam vent tube. An advantage of the manifolding of the steam vent tube is that a natural resistance to hydraulic oscillations is present in the arrangement. Lower frequency natural resonant oscillations within the surrounding two phase region of the fuel bundle are not in phase with higher frequency natural resonant oscillations within the steam vent tube. The result is that the two different resonant frequency systems tend to dampen one another. The tendency of the fuel bundle to resonate at one frequency is reduced. |
062529226 | abstract | A cask is lifted from outside of a nuclear reactor building through an opening which is opened at a portion of the nuclear reactor positioned just above a nuclear reactor pressure vessel. By maintaining the cask under a lifted condition a shroud which is an internal structure of the nuclear reactor is transferred and inserted to the cask from the nuclear reactor vessel. After that, an inlet port of the cask is closed, and the cask together with the shroud is lifted outside of the nuclear reactor building through the opening. The internal structure of the nuclear reactor can be speedily carried from the nuclear reactor building with reduced exposure of operators to radioactivity from the radioactive internal structure. |
054141958 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS The present invention is directed to an on-line system for monitoring, and if desired, control of a particulate material washing processing which produces as a product of remediation a slurry. The particulate material typically would be soil, and the invention will be described as applied to a soil washing process. However, the invention may be applied to processes for remediation of other types of particulate material such as sludges, sediments, scrap yard dust and the like. A typical soil washing process is illustrated in FIG. 1. This process is described in detail in commonly owned U.S. Patent Application Ser. No. 07/529,092 filed on May 25, 1990 which is hereby incorporated by reference into this application. The process is designed to remove from the soil heavy metals and organics. Heavy metals removed by the process include, but are not limited to, cadmium, copper, lead, mercury, and radioactive species such as radium, uranium and thorium. Prior to the process illustrated in FIG. 1, excavated soil is processed to remove large rocks and debris. The soil is then processed in a mechanical size separator 10 to sort and prewash the feed soil with a leachate provided through a line 12. Large pieces of soil, for instance larger than 5 mm, are washed with a leachate, rinsed with water supplied through line 14, checked for residual contaminants, and returned to the site as recovered soil. The effluent of soil particles smaller than 5 mm and the leachate solution discharged from the mechanical separator 10 through line 16 is then processed in a counter-current flow size separator such as the mineral jig 18. In the jig 18, additional leachate solution supplied through line 13 flows upwardly counter-current to the effluent. The fines are carried upwardly with the upward flow of the leachate solution to form a slurry which is discharged through a line 20. These fines typically include heavy metal particles. The velocity of the upward flow of contaminant containing solution in the mineral jig 18 is set to separate fines of a desired size, for example fines smaller than 60 microns in diameter. The slurry discharged in the line 20 includes, in addition to the fines, leachate solution which contains solubilized heavy metals. The intermediate sized particles between 5 mm and 60 microns in diameter, which are discharged from the bottom of the mineral jig 18, are abraded in an attrition scrubber 22 which dislodges mineral slime or fines from them. Intermediate sized particles and the dislodged fines discharged from the attrition scrubber 22 through line 24 are rinsed in a second counter-current flow size separator such as the second mineral jig 26 operated in the manner discussed above in connection with the jig 18. The counter-current flow in the second mineral jig 26 is wash water which flows upwardly at a velocity, again selected to separate the dislodged fines, typically of 60 microns in diameter and smaller. This waste slurry of fines and washwater is discharged through line 28. The remaining intermediate sized particles discharged from the second mineral jig 26 are processed in a density separator such as a cross-current flow jig 30 to extract higher density heavy metal solid waste particles. The cross-current flow carrying the intermediate sized soil particles is discharged through a line 32 into dewatering apparatus such as for instance a clarifier 34 or a hydroclone. Sludge from the clarifier 34 is pumped by a pump 36 onto a drying pad 38. The dried particles recovered from the drying pad are checked for cleanliness and returned to the site as additional clean soil. Water removed from the clarifier 34 is circulated by a pump 40 through a line 42 as the counter-current washwater for the second mineral jig. 26, and through line 44 as the cross-current flow for the density separator jig 30. The two waste slurry streams in the lines 20 and 28 from the first and second mineral jigs 18 and 26, respectively, are discharged into a waste slurry treatment system 46 which precipitates out the dissolved heavy metals for removal with or separately from the fines for disposal or further treatment, and recirculates the leachate through line 48 to makeup tanks 50 for recirculation by pumps 52. The soil washing process must be operated to meet the remediation levels required by environmental agencies. It is desirable to analyze the soil slurry before the costly dewatering step, to determine if the process is meeting these standards. Thus, in accordance with the invention, a monitor device 54 downstream of the mineral jig 26, analyzes the slurry of intermediate sized particles and washwater to determine whether environmental standards on the concentration of heavy metals has been met. If not, a process controller 56 responds to the monitor and adjusts the counter-current flow in the jig 26. For instance, if the heavy metal concentration in the particulate fraction of the slurry exceeds the set point value, the counter-current flow in the jig 26 is increased to divert more of the fines which typically contain the heavy metals, into the waste slurry line 28, thereby reducing the concentration of heavy metal in the slurry discharged by the jig 26. The process controller 56 could also modify the operation of the mineral jig 18 if desired. Furthermore, a monitor 54 could be located at other points in the process in place of, or in addition to, the location shown in FIG. 1. One embodiment of the monitor 54 is the x-ray spectrometer system 58 illustrated in FIG. 2. The spectrometer system 58 includes a sensor module 60 which contains a flow cell 62 through which a stream of slurry 64 diverted from the main flow of the soil washing process by conduit 66 is caused to flow past a window 68 which is transparent to x-rays. An x-ray tube 70 powered by a high voltage supply 72 directs a beam of x-rays 74 through the window 68 into the slurry 64. The x-rays 74 cause secondary emission or x-ray fluorescence of the heavier elements in the slurry 64 including heavy metal contaminants. These secondary emission x-rays 76 are detected by a detector 78 which generates an electrical signal proportional to the amplitude of the secondary x-rays which is amplified in a preamplifier 80. A control module 82 includes a system control 84 which controls the high voltage power supply 72. Module 82 also includes a pulse processor 86 which processes the signal generated by the preamp 80 and analog to digital converter 88 which digitizes the processed detector signal for input to a data processor 90. The data processor analyzes the detector signal to determine the concentration of contaminants in the slurry 64. The data processor 90 also controls the system control 84. A key feature of the invention is the determination of when the soil washing process is running out of specifications. The x-ray fluorescence spectrometer system 58 measures the magnitude of the secondary emission of x-rays of energy levels characteristic of the contaminant of interest. The slurry though is not a representative sample of the desired product. Rather, the soil in the slurry is the desired sample, but it is a variable component of the slurry. The soil in the slurry varies in at least two ways. First, the solids content of the slurry varies. In a soil washing process, the soil content can vary from about 5% to 40% by volume or more, although for a particular soil washing process for treating soil from a particular site, the solids content will remain fairly constant at a specific location in the process. In addition to the variations in the solids content of the slurry, the soil can also vary in chemical composition. In order to determine the concentration of the contaminant, i.e., the ppm, it is necessary to also have a measure of the solids content of the sample being irradiated. In accordance with the invention, the magnitude of the line in the energy spectrum of the secondary emissions of the characteristic energy of the contaminant of interest is measured to determine the amount of the contaminant present. The solids content is measured in accordance with one aspect of the invention by measuring the Compton scatter adjacent the characteristic energy of the contaminant of interest. Water in the slurry generates little scatter, thus, the solids content can be measured as a function of the Compton scatter. The characteristic wavelength of the contaminant of interest and the Compton scatter can be measured using an energy dispersive x-ray spectrometer as the detector 78. The magnitude of the background scatter adjacent the characteristic energy of the contaminant of interest is easily measured in this instance. Alternatively, a wavelength dispersive x-ray spectrometer could be used for this purpose. In this case, a detector for the wavelength of the contaminant of interest is used and a separate detector for a wavelength adjacent the wavelength of the contaminant, but not the wavelength of any element present in the slurry, is used to detect the background. The wavelength dispersive spectrometers provide more accurate measurement but require separate detectors for each of the wavelengths to be measured which makes them less flexible and more expensive for analyzing soils from different sources for different contaminants. The x-ray spectrometer system 58 is calibrated for a particular site using a number of samples with varying concentrations of the contaminant and varying solids contents. Once calibrated, the system can be used on-line to monitor the soil washing process in the manner described above. In accordance with another aspect of the invention, the solids content of the slurry is measured by measuring the amount present in the slurry of selected elements which are present substantially homogeneously in the soil being processed. In accordance with this approach, a number of samples of the slurry with varying levels of contamination and varying solids contents are analyzed for the contaminant of interest and several elements likely to be present more or less homogeneously throughout the soil. The data from these test runs are analyzed using mathematical techniques such as partial least squares and the method of principal component regression, which are well known in the field of analytical chemistry, to determine which lines are representative of the amount of soil present in the slurry. Such representative elements would be dependent on the particular soil as determined from these tests, but exemplary constituents could be: rare earths, silicon, iron and aluminum. In another embodiment of the invention, the data processor 90 includes a neural network which applies a pattern-recognition approach to determining the concentration of contaminants in the soil present in the slurry. In such an approach, the neural network is exposed to a training set of samples that represent the variation of elemental composition that is expected to be observed during the monitoring period. The neural network is also "trained" by being given the information that the contaminant is present at a concentration that is above or below the desired remediation level. In this way, the neural network "learns" to interpret the x-ray fluorescence spectra of contaminated soils, after cleaning, and to report whether the process is within specifications or out of specifications. This is the minimum information necessary for soil cleaning process control. The advantages of the pattern recognition approach are that the neural network may be tolerant of unexpected deviations in soil composition that were not represented in the training set, and that the neural network may perform adequately with minimal information that would be inadequate for a more mathematically rigorous calibration technique. Another aspect of the invention is directed to the window 68 in the flow cell 62 of the x-ray spectrometer system 8. This window is subject to abrasion by the soil particles in the slurry 64. We have found that a thin diamond film 69 applied to the inner surface of the window 68 or even free-standing diamond film windows provide suitable erosion resistance. At the same time, the polycrystalline diamond, being carbon with an atomic number of 6, is substantially transparent to x-rays. Commonly owned U.S. Patent application Ser. No. 07/600,952, filed on Oct. 22, 1990 and entitled "Diamond Coated Optical Window and Methods of Making Same," now abandoned, discloses diamond coated windows in which a polycrystalline diamond film is deposited on a substrate. The free surface of the film is then bonded to the window material using a chalcogenide glass. The substrate is then removed. Suitable chalcogenide glasses for securing an erosion resistant layer to an optical window are disclosed in U.S. Pat. No. 4,072,782. Another embodiment of the invention is based on prompt neutron activation (PNA). The basis of the PNA approach is the thermal neutron capture reaction which is expressed symbolically by the following: M.sub.X (n,.gamma.).sup.M+1 X. In this reaction, a nucleus of the element X with a mass of M captures a thermal neutron transmuting it to another isotope of X with a mass M+1. The latter nucleus is in an excited state when it is formed and it will de-excite promptly (<10.sup.-6 sec) to the ground state by emission of gamma rays. This can be seen more clearly from the energy level diagrams of FIG. 3. These diagrams show the transition after neutron capture to an excited state in the isotope .sup.M+1 X from which the de-excitation proceeds directly to the ground state with the emission of a gamma ray with energy E.sub.1 or through intermediate levels which result in gamma rays with energies E.sub.2 and E.sub.3. The probability of the emission of these gamma rays, or branching ratio, is well known so that the coupling of their energies and rates of emission provide signatures of the isotope .sup.M+1 X and also, via the reaction which produced it, of the isotope .sup.M X. The activation rate A, i.e., the number of these reactions per second, is governed by the equation: EQU A=n.sigma..phi. Eq. 1 where n is the number of target nuclei, .sigma. is the neutron capture cross section in cm.sup.2, and PA1 .phi.is the thermal neutron flux in n/cm.sup.2 /sec. PA1 The two parameters which can be controlled in this equation are n and .phi., so that, basically, the PNA system should be designed to irradiate a large n with a high .phi.. The emission rate of gamma rays is determined by multiplying the activation rate A by B, the branching ratio for a particular gamma ray. PA1 B is the branching ratio (unitless) PA1 .alpha. is the attenuation factor (unitless) PA1 E is the efficiency of a gamma ray detector, and PA1 G is the solid angle presented by the detector array. The complete equation for the system must factor in the gamma ray attenuation in the material between its point of origin and a gamma ray detector, the solid angle of the detection system and the efficiency of the detector for producing a full energy response to the intercepted gamma ray and may be stated as follows: EQU C=n.sigma..phi.B.alpha.EG Eq. 2 Where C is the counting rate in counts/sec There are three basic elements required for PNA system operation: a thermal neutron field for the irradiation, a means for transporting the material into and out of the thermal neutron field, and a detection system for detecting the gamma ray response of the material to the thermal neutron field. Shielding, although not directly related to the system's operation, is required as well. Thermal neutron fields for activation analysis are produced by moderating fast neutrons from a source to thermal energies. Basically, there are two types of sources: machine and isotope. Machines are available which produce neutrons of about 3 MeV. The main advantages of machines are that they produce no background gamma ray radiation, per se, and they can be turned off. Their disadvantages are that they are bulky, consume large amounts of power and are expensive (initial expense plus cost of replacement tubes). Thus, except for limited applications, an isotopic source is preferred. The leading candidate is .sup.252 Cf which produces a fission distribution of neutrons at the rate of 2.34.times.10.sup.12 n/sec/gm. Because of this high rate one can obtain a significant neutron emission rate for a small quantity of material. Typically, they are encapsulated in a stainless steel or zinc alloy cylinder with the dimensions, in. diam. and about 1.5 in. length. The small size makes it particularly suitable for field applications where bulk and weight must be kept at a minimum. Two disadvantages are that .sup.252 Cf is also an intense source of gamma rays and that it cannot be turned off. To obtain the thermal neutron field, the high energy source neutrons must be moderated to thermal energies, i.e., energies around 0.025-0.04 eV. At these energies, a neutron is as likely to gain kinetic energy as to lose it in a collision with a nucleus. As well, at these energies, the neutron capture cross section is high, thereby enhancing the probability of a capture reaction. Moderation is typically accomplished by surrounding the source with hydrogenous material, e.g., water, polyethylene, parraffin, etc. However, in this instance where the sample material is a waterbased slurry and a determination must be made of the solids content of the slurry, the introduction of additional hydrogenous material is to be avoided. Instead, deuterium which has a very small capture cross section is the preferred moderator. A spectrometer system employing prompt neutron activation (PNA) system 92 is shown in FIGS. 4 and 5. In PNA system 92, a sample of the slurry is delivered through a conduit 94 to a rectangular measurement chamber 96. The chamber 96 is preferably oriented for vertical flow of the slurry 98 to prevent build-up within the chamber. A pair of .sup.252 Cf sources 100 are spaced along one side of the measurement chamber 96 and surrounded by deuterium moderator material 102. A series of gamma ray detector elements 104 are mounted around the measurement chamber 96. Preferably, reverse electrode germanium detectors with energy resolution of a few KeV for MeV gamma ray energies are used. Protection should be provided from background gamma ray radiation. One source of such background radiation is the .sup.252 Cf neutron source 100. Accordingly, the source 100 is encapsulated in a gamma ray shield 108. Virtually any heavy metal could be used as a shield, however, bismuth is superior because of its very small neutron capture cross section and its relatively low energy gamma ray emission. Another technique would be to position the detector assemblies 104 such that they do not "face" the neutron sources 100. The detector assemblies 104 further include bismuth shielding 110 around the annular detector elements 106 which will further reduce the .sup.252 Cf background, as well as any moderator produced background. In order to further diminish the background and enhance detector sensitivity, an annular gamma ray detector 112 is provided between the primary gamma ray detector 106 and the bismuth shield 110. The counts generated by the individual detector assemblies 104 are transmitted to a data processor 114. The data processor 114 can employ anti-coincidence techniques to eliminate pulses occurring simultaneously in the annular detectors 112 and primary detectors 106 due to gamma rays entering from the side, such as from the moderator. Furthermore, this technique would also eliminate single and double escape pulses, thereby cleaning up the spectrum. The data processor 114 processes the signals from the detectors 104 to determine the number of atoms of the contaminant of interest in the measurement volume. Since the hydrogen atoms in the water of the slurry also emit gamma rays of a characteristic energy, the number of hydrogen atoms, and therefore the amount of water in the measurement volume can also be calculated by the data processor 114. By utilizing a density measurement, taken for instance by a weighing device 116 in the supply conduit 94, the solids content of the slurry can be calculated. With the solids content, and the amount of residual contamination detected, the concentration of the residual contamination can be determined. An evaluation of the system disclosed in FIGS. 4 and 5 can only be made with any accuracy when the type of waste material and the potentially hazardous element composition had been identified. For purposes of illustration, it will be assumed that the soil is to be examined for traces of heavy metals, for instance, U, Th, Cd, Cu, and Pb. The U.S. EPA guideline limits for concentrations of Cd, Cu and Pb in soil are 3, 170 and 50 ppm, respectively. The limits for U and Th are given in activity concentration and are 30 and 23 pCi/gm, respectively. A review of the neutron capture cross sections of the elements mentioned above, and their concentrations indicate that Pb will be among the most difficult to detect. The discussion will focus on that element and present the results for the others in a table. The goal in these calculations is to determine whether the PNA system disclosed in FIGS. 4 and 5 can detect the metals at the concentration levels mentioned above. It is assumed that the soil is a slurry with a particulate concentration of 15% by volume. It is further assumed that the measurement chamber 96 is a rectangular parallelpiped with a height of 15 cm, width 30 cm and a length of 60 cm in the direction of flow. The .sup.252 Cf sources 100 are 800 .mu.gm each. This will produce an average thermal flux of about 10.sup.7 n/sec throughout the volume of the measurement chamber 96. Twenty germanium gamma ray detectors 106, each 65 mm in diameter are deployed about the volume 96. Using the count rate Equation 2 above, it is estimated that a 50 ppm concentration of Pb will give about 2-3 cts/sec. of 7.4 meV gamma rays. This is slightly low for determination of whether the soil would meet U.S. EPA guidelines; however, this approach would easily meet New Jersey's 400-1000 ppm and California's 1000 ppm guidelines. It is estimated that with small modification of the source strength or configuration that the 50 ppm level, the U.S. EPA level can be made to produce an acceptable gamma ray count rate taken to be about 10 cts/sec or more. Table 1 presents the detection limits of the five metals mentioned above. It is arbitrarily been assumed that approximately 10 cts/sec. as a reasonable counting rate to obtain the sensitivity limit indicated. The table indicates that the system of FIGS. 4 and 5 is expected to meet the U.S. EPA guidelines for Pb and U with the system operational parameters given. For other elements listed, the detection sensitivities are far superior to the requirement of the U.S. EPA guidelines. TABLE 1 ______________________________________ Sensitivity.sup.+ of Proposed PNA System to 5 Hazardous Elements U.S. EPA Soil Concentration Expected Element Element PNA Sensitivity ______________________________________ Pb 50 ppm .apprxeq.50 ppm U 30 pCi/gm 30 pCi/gm Th 23 pCi/gm 5 pCi/gm Cd 3 ppm 0.35 ppm Cu 170 ppm 8.21 ppm ______________________________________ .sup.+ Assumes soil to be 15% of the slurry volume; dry soil sensitivitie should be better than those listed. The determination of the amount of particulate material present in the slurry can alternatively be measured by measuring the magnitude of gamma rays emitted by constituents which are present in the particulate material substantially homogeneously, as discussed above in connection with x-ray fluorescence. The prompt neutron activation system provides a determination of the residual contaminant concentration for the entire volume within the measurement chamber, while the x-ray fluorescence technique only examines the sample to a depth of about 100 microns to about 1 mm. As long as the x-ray fluorescence sample is representative of the slurry it can provide suitably accurate concentration calculations. Another embodiment of the invention is illustrated in FIG. 6. This embodiment utilizes laser induced breakdown (LIB) with spectroscopic analysis. A sample of the slurry 118 provided through conduit 120 is passed through a flow cell 122 having an optical window 124. A high power laser 126 generates a laser beam 128 which is guided by deflecting mirrors 130 (or a light guide), and focused by a lens system 132 through the window 124 onto the slurry 118. The high electric field of the focused laser pulse produces a bright hot spark through dielectric breakdown of the slurry. Light 134 from the laser induced spark 136 is collected by a lens 136 and focused onto the end of an optical fiber 138 which transmits the light to the entrance slit of a spectrometer 140. The spectrometer 140 measures the intensity of light characteristic of the contamination of interest. In order to determine the concentration of the residual contaminant, the spectrometer 140 also measures the intensity of light characteristic of elements which are ubiquitous in the soil as a measure the amount of soil vaporized in the laser beam. The selection of these reference elements and the concentration of the residual contaminant can be performed using the techniques described above in connection with the x-ray fluorescence spectroscopy. The spectrometer 140 provides the measurement of the intensity of the selected light energy to a data processor 142 which calculates the concentration of residual contaminant in the soil. A system control 144 controls the laser 126, spectrometer 140 and data processor 142. The monitoring system illustrated in FIG. 6 will generate large amounts of data at a high rate. Consider a laser firing at 100 shots per second, and a detector with 1000 channels of intensity data at 12 bit resolution. The minimum data rate to handle such data would require a capacity for 300 kbytes/sec. The requirement for calculating averages and concentrations places even greater demands on the analysis system. Thus, a workstation type computer is used as the data processor 142. 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 arrangements 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. |
abstract | A combination of a top nozzle and a guide thimble of a nuclear fuel assembly and, more particularly, a structure for joining an inner-extension tube, the top nozzle and the guide thimble. When an inner-extension tube head, which is provided as a means for facilitating removal of the top nozzle of the nuclear fuel assembly from the guide thimble, is removed from an inner-extension tube body to separate the top nozzle from the nuclear fuel assembly, the inner-extension tube body is prevented from undesirably rotating, so that the guide thimble and the inner-extension tube body can maintain the joined state. |
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048779628 | summary | BACKGROUND OF THE INVENTION This invention relates to a method of performing ion implantation on a silicon substrate, and more particularly, it relates to an ion implantation method by means of which planar channeling in the substrate can be reduced. Ion implantation is widely used in the manufacture of semiconductors in order to introduce impurities into silicon substrates on account of the superior reproducibility, uniformity, and controllability which can be achieved. FIG. 1 schematically illustrates a conventional electrostatic scanning ion implantation apparatus by means of which ion implantation is commonly carried out. As shown in this figure, ions are emitted from an ion source 1. Of these ions, undesired ones are removed by an analysis magnet 2. An ion beam 6 containing the desired dopant ions exits from the analysis magnet 2 and passes through Y-scan electrode plates 3 and X-scan elecrode plates 4, which control the direction of the ion beam 6. The ion beam 6 is directed at a silicon wafer 7 which is supported by a base 5. In order to reduce axial channeling of the dopant ions, the wafer 7 is usually tilted by an angle T so that the angle of incidence of the ion beam 6 with respect to a normal to the surface of the wafer 7 is about 7.degree.. FIG. 2 is a sheet resistance map of the surface of a silicon wafer 7 which was subjected to ion implantation by a conventional method using the apparatus of FIG. 1. Sheet resistance was determined by the four-probe method of resistivity measurement. The wafer 7 was made of (100) Si having a wafer flat 7a lying in a (110) crystal plane. The angle of tilt was 7.degree., and the angle of the wafer flat 7a with respect to the horizontal was 0.degree.. Implantation of boron ions was performed at 50 keV at a dose of 2.times.10.sup.13 atoms per square cm, after which annealing was performed at 900.degree. C. The two bold contour lines in FIG. 2 are reference values, and the other contour lines are drawn at intervals of 0.5% of the reference values. The (+) marks indicate regions where the sheet resistance was larger than the reference value, and the (-) marks indicate regions where the sheet resistance was smaller than the reference values. As is clear from FIG. 2, there is a belt-shaped region of low sheet resistance at the center of the wafer 7, and the sheet resistance increases towards the top and bottom of the wafer 7. The cause of this decreased sheet resistance is not an increased dose of dopants at the center of the wafer 7. Rather, it is due to planar channeling. Namely, even though axial channeling along the <100> crystal axes was prevented by tilting the substrate by 7.degree. with respect to the ion beam 6, as the wafer flat 7a was parallel to the horizontal, planar channeling took place between the (110) crystal planes, and dopant ions penetrated deeper into the surface of the wafer 7 than desired, just as when axial channeling occurs. When using an electrostatic scanning ion implantation apparatus in which an ion beam is scanned in the X and Y directions, the ion beam can easily become aligned with the (110) crystal planes at the center of the wafer 7, so that in the center of the wafer 7, the depth of implantation increases and the sheet resistance decreases. SUMMARY OF THE INVENTION Accordingly, it is an object of the present invention to provide a method for ion implantation of silicon substrates which can prevent both axial and planar channeling, thereby increasing the uniformity of doping of the substrate. In the present invention, a substrate of (100) silicon is tilted with respect to an incident ion beam to prevent axial channeling, and at the same time, the substrate is rotated in its own plane to a position such that the (110) crystal planes of the substrate are not aligned with the incident ion beam. Ion implantation is performed four times on a single substrate at four different angles of rotation which differ from each other by multiples of 90.degree., using one-fourth of the total desired dose of ions each time. As a result, the (110) crystal planes are never aligned with the ion beam, and planar channeling is prevented. An ion implantation method in accordance with the present invention comprises: positioning a substrate of (100) Si in a first position such that it is tilted with respect to an ion beam with which it is to be irradiated and such that it is rotated in its own plane by a rotational angle of 15.degree. to 75.degree. from a position in which the (110) crystal planes of the substrate would be aligned with the ion beam; irradiating the substrate with the ion beam in the first position using a dose of ions equal to approximately 1/4 of the total desired dose of ions to be implanted into the substrate; rotating the substrate in its plane to a second position which is 90.degree. removed from the first position and irradiating the substrate with the ion beam using the same dose as before, approximately equal to 1/4 of the total dose; rotating the substrate in its plane to a third position which is 180.degree. removed from the first position and irradiating the substrate with the ion beam using the same dose, i.e., approximately equal to 1/4 of the total dose; and rotating the substrate in its plane to a fourth position which is 270.degree. removed from the first position and irradiating the substrate with the ion beam using the same dose, i.e., approximately equal to 1/4 of the total dose. The rotational angle of the substrate when in the first position is preferably about 45.degree. from a position in which the (110) crystal planes would be aligned with the ion beam. The angle of tilt of the substrate is preferably about 7.degree.. In a preferred embodiment, the present invention is carried out using an electrostatic scanning ion implantation apparatus, but any type of ion implantation apparatus can be employed, provided that it is equipped with a base for supporting a substrate which is able to tilt the substrate with respect to an incident ion beam and is able to rotate the substrate in its own plane. |
041982720 | claims | 1. In a liquid metal cooled fast breeder nuclear reactor having a fuel assembly support structure, and a peripherally restrained up-standing fuel sub-assembly, the improvement wherein said sub-assembly comprises: a bundle of spaced nuclear fuel pins; a tubular wrapper enclosing said bundle of fuel pins; a spike extension for the tubular wrapper for plugging the fuel sub-assembly into said fuel assembly support structure, said spike extension comprising a sleeve having axially spaced cylindrical surfaces defining spigots for engaging vertically spaced sockets in the fuel assembly support structure; said sleeve being spaced from said wrapper and having an end region enclosed by an end region of the wrapper to define an annular clearance therebetween; and an elongate resilient tie member pivotably connecting the tubular wrapper and sleeve, the tie-member extending through the sleeve and having one end rigidly secured to the wrapper and another end rigidly secured to the sleeve. 2. An improved nuclear reactor according to claim 1 wherein the sleeve carries resilient split rings for sealing against liquid metal cooling flow between the wrapper and sleeve. |
060842431 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS Referring now to the drawings for the purposes of illustrating the present preferred embodiments of the invention only and not for the purposes of limiting the same, the Figures show a radiation absorption gauging system, generally shown as 10, installed on a conduit or vessel 12. The radiation absorption gauging system 10 includes a radioactive source, generally shown as 14, and a detector, generally shown as 16. The radioactive source assembly 14 and the detector 16 are attached on opposite sides of the conduit 12 by means of a mounting assembly generally shown as 18. The mounting assembly 18 includes a first mounting bracket 20 to which the radiation source 14 is attached and a second mounting bracket 22 to which the detector 16 is attached. Threaded rod and bolts 24 connect the first and second mounting brackets 20 and 22, respectively. The detector 16 includes a scintillation detector 26 which is used to measure the strength of the radiation field after it has passed through the pipe 12 and any process material therein. The strength of the signal is related to the density of the process material. As the density of the process material increases, the strength of the signal decreases. The scintillation detector includes a scintillation crystal, generally shown as 28, and a photomultiplier tube, associated electronics and voltage divider, generally shown as 30. A lead collimator 32 is positioned around the end of the scintillation detector 26. The collimator 32 functions to shield the scintillation detector from stray background radiation which will affect the measurement. A microprocessor amplifier 33 is coupled to the detector 26 to calculate the desired measurement. In use, the radioactive source 14, such as cesium-137, generates gamma waves. Alternative sources include cobalt-60 or americium-124. The gamma waves exit the radioactive source 14 from the side facing the conduit 12. The gamma waves travel through he conduit 12 where some of it is absorbed by the material therein. The gamma waves strike the scintillation crystal 28 generating a light photon. The detector 16 converts the light photon into digital pulses which are transmitted to the microprocessor 33 which is calibrated to transform the digital pulses into representative units such as density or concentration. The radiation source 14 includes one of more source holders generally shown as 40 having a plurality of radioactive sources 42 mounted therein. More specifically, the source holder 40 includes a cylindrical housing body 44 formed from a lead alloy, tungsten alloy, a steel alloy, or other high density material (possibly even spent uranium) and having a central cavity or bore 46 passing therethrough. By means of example, the housing body may be approximately 3.5 to 4.5 inches in diameter, 0.24 inches high and the bore 46 may be 0.96 inches in diameter. The radiation sources 42 are, for example, thin disks with less than 10 micro-curies of cesium-137 deposited on 9 milligram per square centimeter of polyamide and covered with a 0.010 inch of aluminized mylar foil. The cesium-137 serves to generate gamma rays for the detection purposes described herein. Alternatively, the sources 42 may be formed as small spheres, pellets, rods, etc. and may be formed from cobalt-60 or americium-241. The diameter of the bore 46 is approximately equal to the outer diameter of the disks on which the radioactive sources 42 are mounted, in such embodiment. Inert filler plugs 47 or tape are provided on the outer ends of the stack of radioactive sources 42. It is important to note that the thickness of the source holder 40 is sufficient to prevent radiation exceeding a predetermined level, such as 0.5 milliRem per hour or 0.1 milliRem per hour in some cases, from being present on the outer radial surface of the source holder 40. Indeed, it is a principal feature of this invention that in each embodiment disclosed, be it of the system 10, radiation source 14, the source holder 40 or the shipping container 60 (described below), and other embodiments thereof, that the level of radiation on any surface other than in the radiation beam path be below a predetermined limit. Typically, such predetermined limit will be that level which may be handled and shipped without significant regulation. Currently, such levels are less than 0.5 milliRem per hour in most situations and less than 0.1 milliRem per hour in some cases. In accordance with this invention, a plurality of the radioactive sources 42 are permanently preassembled into the source housing body 44. As used herein, "permanently" is intended to indicate that the radioactive sources 42 cannot be removed from the housing body without damaging the sources 42, the holder assembly 40 or the retainer therefor. "Preassembled" is intended to mean that the radioactive sources 42 are permanently fixed in the housing body 44 at a time and place remote from the installation of the radioactive sources into the radiation absorption gauging system 10. As shown in FIG. 1 one or more radiation source holder assemblies 40 may be mounted on the first mounting bracket 20 by means of a mounting clamp adapter plate 48 as shown in FIG. 1. A front support plate 50 is provided between the source holder 40 and the mounting clamp adapter plate 48. A cylindrical lead end cap 52 is provided on the side of the radiation source 14 remote from the mounting clamp adapter plate 48. By means of example, the end cap 52 may be 1.4 inches in height and 3.5 to 4.5 inches in diameter. A rear support plate 54 is provided adjacent to the end cap 50. One way fasteners, such as screws 56, are provided to clamp the entire assembly including the rear support plate 54, end cap 52, radioactive source holders 40 and the front support plate 50 to the mounting clamp adapter plate 48 so that removal therefrom is difficult. In addition, a lead storage cap 58 may be secured to the end cap 52 by means of screws 60. For example, the storage cap 58 may be 1.4 inches in height and 3.5 to 4.5 inches in diameter. Alternatively, the storage cap 48 may be hingedly attached to the mounting clamp adapter plate 48. In any event, in its storage condition, the assembly must provide surface radiation levels less than, for example, 0.5 milliRem per hour. Also, alternatively, the end cap 54 and the source holders 40 may be formed as a unitary component. Indeed, the entire radiation absorption gauging system 10 may be formed as an assembly when conduit 12 is a flanged conduit that can be inserted into another line, so long as the radiation on any external surface of system 10 is less than a predetermined amount, such as 0.5 milliRem per hour. In accordance with a present preferred embodiment of the invention, there is shown a shipping container, generally shown as 60, for the source holder 40. As shown in FIG. 3, the shipping container 60 includes a first container body member 62 and a second container body member 64 which, when assembled, are adjacent the first face 66 and second face 68, respectively, of the source holder 40. Preferably, the first and second housing members 62 and 64, respectively, are formed from a lead material or a lead alloy, tungsten alloy, steel alloy or other high density material and are cylindrically shaped so that their outer diameters are approximately the same as the outer diameter of the source holder 40. For example, the first and second body members 62 and 64 may be 1.4 inches in height and 3.5 to 4.5 inches in diameter. However, the housing members may be of alternative shapes, such as square shaped. Bores 70 are provided through the first and second body members 62 and 64, respectively, and through the source holder 40 so that bolts 72 and nuts 74 may join the assembly in forming a shipping container 60 for shipment of the source holder 40. Alternatively, the first body member 62 may be unitary with the source body housing 44. As used herein, "unitary" means that such members may be formed as a single piece or formed as two pieces that are joined with the joint being sufficient to prevent more radiation from passing therethrough than through the walls of the members. The thickness of the first and second housing members 62 and 64 is sufficiently great that at all points on the periphery thereof, the level of radiation is less than 0.5 milliRem per hour. All components of the shipping container 60 are assembled and shipped in a shipping package, such as a tube, box or other package formed from plastic, cardboard or other suitable material. In use, a user receiving the shipping container 60 may remove nuts 74 from bolts 72 and remove the first body member 62 from the source holder 40. The source holder 40 may then be inserted into the radiation source 14 in an assembly 10 and the first body member 62 reassembled onto bolts 72 by means of nuts 74 and the shipping container returned to the manufacturer. By virtue of the provision of the shipping container 60, the source housings 40 will be shipped in commerce in compliance with applicable regulations. It should also be understood by those skilled in the art that the shipping container 60 may also serve as part of the radiation source assembly 14 wherein the second body member 64 acts as the end cap 52. Also, the storage container 80 may also serve as the shipping container 60. When it is desired to deactivate the system 10, such as for maintenance, the source assembly 14 should be removed from the conduit 12 and placed into a safe condition by forming a storage assembly 80. To accomplish this, the source holder 40 is removed from the first bracket 20. Thereafter, the screw 59 is removed thereby allowing the removal of rear housing 58. Rear housing 58 is then provided on the opposite side of the mounting clamp adapter plate 48 and the screw 59 is used to attach it to the mounting clamp adapter plate 48. As such, the radioactive sources 42 will be surrounded by lead material so that in all directions surrounding the storage assembly 80, the level of radiation is less than, for example, 0.5 milliRem per hour. Another embodiment of the invention is shown in FIG. 7 wherein the elements have similar numerals to those described above but increased by the value of 100. In this embodiment, there is shown a radiation absorption gauging system 100 for use in detecting the presence or absence of process material and provides a minimum or maximum level indication for tanks, pipes, hoppers and chutes. The system 100 includes a first radioactive source 114a and a first detector 116a mounted on opposite sides of the tank 113 to indicate a first level 102. There is also provided a second radioactive source 114b and a second detector 116b to indicate a second level 104. It will be appreciated that the construction of detectors 116a and 116b are similar to that of the detector 16 described above. Furthermore, the construction and operation of the radioactive sources assembly 114a and 114b are similar to those described above in respect to radioactive source assembly 14. The operation of system 100, however, merely serves to indicate whether there is material between the respective radiation sources and detectors. As such, if detector 116b detects material and detector 116a detects material, the system may generate an output indicating that a first high level 102 is achieved in the tank 113. However, if detector 116b detects material and detector 116a does not, the system will generate a signal indicating that the level of material in vessel 113 is between the respective high 102 and low 104 levels. If, however, detector 116b does not detect material in vessel 113, a signal will be generated indicating that the level of material is below that of the low level 104. It will, of course, be appreciated by those skilled in the art that if only a single level measurement were desired, only one source/detector combination such as 114a and 116a or 114b and 116b may be employed. Another embodiment of the invention is shown in FIG. 8 wherein a radiation absorption gauging belt weighing system 410 is disclosed. In the belt weighing system 410, a belt 402 movably supports material to weighed 404 therein. One or more radiation source assembly(s) 414 are provided on one side of the belt 402 and a radiation detector 416 is provided the opposite side with source assemblies 414 either above or below the belt 402. As such, the material 404 passing on the belt 402 absorbs some of the radiation passing from the source 414 to the detector 416. By proper calibration, the weight of the material 404 can be calculated by a microprocessor 433 based on the amount of radiation which is ultimately sensed by the detector 416. Yet another embodiment of the invention is shown in FIG. 9. This embodiment provides a continuous level indication of materials in tanks, pipes, hoppers and chutes, such as vessel 213. In this system, a detector 216 is provided similar to detector 16 described above. However, the radioactive source indicated as 214 is instead arranged in a column. That is, the radioactive source 214 includes a plurality of radioactive sources which are mounted in a housing generally shown as 290. The housing 290 is preferably formed from lead, lead alloy, tungsten alloy or steel alloy. The housing 290 is preferably formed as two members 292 and 294, respectively, which may be joined in end-to-end relationship by fasteners such as screws. The housings 292 and 294 may each be, for example, 6.25 inches by 3 inches by 1.5 inches. The housings 292 and 294 have laterally extending flanges 296 and 298, respectively. Additionally, end members 300 and 302 are attached to the members 292 and 294, respectively, by fasteners such as screws. The end members 300 and 302 may be, for example, 3 inches by 1.5 inches by 1.25 inches. The members 292 and 294 have a series of recesses 304 formed therein in a colurn on their faces 293 and 295, respectively. The recesses may either receive a radioactive source 242 which are like the sources 42 therein to generate an elongated radiation field. As such, the housing 290 may be attached to a vessel 308 by fasteners through flanges 296 and 298 to measure the level of material therein. Such measurement is made possible by virtue of the elongated radiation field created by the sources 242 and the detection and calculation capabilities of the detector 216. Again, it will be appreciated that due to the structure of the housing 290, the level of radiation on all surfaces thereof, except in the direction of the radiation beam path normal to the faces 293 and 295, will be at an acceptably low level such as 0.5 milliRem per hour. The housing 290 may also serve as a shipping container 291 for the radioactive sources 342 mounted therein. In that case, the housings 292 and 294 may be separated and positioned with faces 293 and 295 in confronting relationship. Fasteners then join the flanges 296 and 298 to one another An end protector 309 is provided to cover and join the ends of the housings 292 and 294 remote from end members 300 and 302. The end protector may be 3 inches by 3 inches by 1.5 inches. Again, it is important that, when assembled, the level of radiation on the surface of the shipping container 291 be acceptably low, e.g., 0.5 milliRem per hour. Thus, from the foregoing discussion, it is apparent that the present many of the problems encountered by prior radiation absorption gauging systems are overcome. Those of ordinary skill in the art will, of course, appreciate that various changes in the details, materials and arrangement of parts which have been herein described and illustrated in order to explain the nature of the invention may be made by the skilled artisan within the principle and scope of the invention as expressed in the appended claims. |
055263860 | summary | FIELD OF THE INVENTION The present invention relates generally to improving the efficiency, reliability, safety, steam mixing, and other operational characteristics of nuclear electrical power plants with a light water reactor. And more particularly the instant invention relates to mixing superheated steam from a heat recovery boiler of a gas turbine combined cycle steam system with main steam at the HP stage and the LP stage of the main turbine. BACKGROUND OF THE INVENTION In the operation of nuclear power plants, there are several important problems that significantly increase the cost of operations and maintenance. Some of these significant problems are: inherently low thermal efficiency, erosion-corrosion fuel storage and waste disposal limitations, equipment maintainability and reliability concerns as well as the limited fuel availability and the expensive requirements of storing spent fuel. Various approaches have been tried to solve these problems separately, but few of the existing methods attack more than one of these concerns at a time. Currently, in nuclear power plants (both pressurized water reactors or boiling water reactors), steam at the inlet of the turbine is saturated. This results in lower turbine efficiency and intensive erosion-corrosion problems. There have been several attempts to use single or multiple fossil reheaters to treat the steam of a operating nuclear power plant. However, while higher steam temperatures due to fossil heating between the low pressure turbine stage and the high pressure turbine stage have been achieved, higher efficiency still has not been demonstrated. One such attempt is U.S. Pat. No. 5,361,377 to John A. Miller. The Miller heat exchanger method and apparatus of treating the steam claims an increase in efficiency by use of a fossil fired reheater. A problem with this method is that the fossil reheater still consumes fuel in a process that is inefficient in itself. In other words a big part of the heat from the burning of the fossil fuel goes up the stack. A process is needed that produces electricity, and utilizes the waste heat from that process to treat the steam in an efficient manner. Mixing of the steam going from the LP stage to HP stage of the main turbine with high enthalpy steam is a solution that brings all of these advantages. It is generally recognized that for mixing to be done in non-isothermal fluids, in boundary flow conditions, a pipe length of about 50 diameters is required. In a typical nuclear power plant the typical main steam pipe outside diameter is approximately 0.8 meter. This would require a pipe of 40 to 50 meters in length to achieve through mixing. Therefore, a more efficient means of mixing is needed to mix the steam from a heat recovery boiler with main steam. It is therefore an object of the instant invention to treat the steam prior to entering the HP stage of the main turbine, then again before entering the LP stage by mixing the nuclear steam with high enthalpy steam from an efficient process. SUMMARY OF THE INVENTION It is therefore an object of this invention to provide a means to improve the efficiency, availability, safety, and steam mixing, and economy of nuclear power generation systems. And more particularly an improved means for providing superheated steam to the steam turbine driving the electrical generator. It is a more particular object of the invention to allow for lower hot leg temperatures (T.sub.hot) in the primary cycle of an existing nuclear power plant. The lower T.sub.hot would reduce the rate of erosion-corrosion and extend the lifetime of PWR steam generators. It is a further object of the invention to extend plant life by enabling the delivery of the same amount of electricity for a smaller amount of heat generated by nuclear reaction. In other words, the invention is a combined nuclear-gas turbine cycle that can be used to, inter alia, reduce corrosion damage to the nuclear steam generator by operating the outlet reactor hot leg temperature below the critical corrosion temperature. Corrosion is significantly reduced by operating steam generators at less than 600.degree. F. (315.degree. C.). With the instant invention nuclear power plants can operate below 600.degree. F. (315.degree. C.) without derating the electrical output of the station. The advantage of reduced corrosion is due to reduced steam generator temperature caused by an increased thermal efficiency of the resulting steam cycle. One factor that causes an increase in efficiency is that the additional electrical output from the gas turbines further adds to the electrical grid output of the station. In the instance where the hot leg temperature is reduced below the critical steam generator corrosion temperature, the thermal power of the reactor will be lower, resulting in a potentially longer operating cycle time between reactor refueling cycles. This enables a longer period between refueling which is getting progressively more expensive. The disposal of spent nuclear fuel and waste, which is now a major concern of nuclear industry, will be reduced significantly. In addition, lower reactor power avails opportunities to perform maintenance on many key apparatuses in the secondary system (the steam cycle in PWRs and the coolant cycle in BWRs) while the reactor is at power. By reducing corrosion and erosion the down time is decreased thus enabling the plant to operate with less down time. It is well known in the art that the fuel replacement cost during downtime is from several hundreds of thousands of dollars to one million dollars for each day, depending upon the plant. When the reactor is operating at a lower thermal power, there are, in essence, built-in spare components with the arrangement of currently installed equipment. To underscore the importance of erosion and corrosion in steam equipment it should be noted that maintenance of many secondary-side (i.e. , steam handling) components are often the source of unplanned outages. In the instant invention, the steam turbine has higher efficiency and reduced erosion-corrosion. The reactor steam, prior to entering the high pressure stage of the main turbine, is superheated by a high-pressure-mixer-superheater (HPMS) that utilizes the highly superheated steam from the heat recovery boiler (HRB) of the gas turbine combined cycle. The steam is then conditioned, according to need, by several different components. In the most commonly used application to improve the efficiency of an operating nuclear power plant the steam would next go to the moisture separator reheater (MSR), that is a common component of existing plants. The MSR can be bypassed depending upon the existing steam quality from the high pressure (HP) stage of the main turbine. The steam is next improved for use by the low pressure (LP) stage of the main turbine. When required this is done in a low pressure mixer superheater (LPMS) which mixes steam from the MSR (when used) with HRB steam. The instant invention has a marked increase in the portion of the LP stage of the main turbine that is driven by superheated steam. This enables a reduction in corrosion and erosion. The invention also provides an additional avenue to make nuclear stations safer due to the addition of an operating source of alternative emergency power (i.e., the gas turbines). Additionally, the size of the gas turbine and generator is not limited and the instant invention therefore readily lends itself to operation in conjunction with a large gas turbine and generator that could function as an independent source of power during peak times or alone to supply a grid. |
058728252 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS Referring now to the figures of the drawing in detail, wherein like parts bear the same reference symbols in both figures, and first, particularly, to FIG. 1 thereof, there is seen an apparatus 1 for inerting and venting a containment atmosphere CA in a nuclear power station. The latter is not illustrated or described in more detail for purposes of brevity of this specification. The apparatus 1 comprises a line 2, into which an activity holdup device 4 is inserted. The line 2, capable of being shut off by means of a valve configuration 5, communicates, on the one hand, via a leadthrough 6 with the interior of a reactor safety vessel or containment 8 of the nuclear power station and, on the other hand, with a distributor 10 in the form of a threeway valve. In the distributor 10, the line 2 branches into a supply line 12 for an inerting agent I and into a vent line 14 for containment atmosphere CA. The supply line 12 has a heat exchanger 16 (superheater) provided with a control device 15 for temperature regulation inserted. The supply line 12 is connected to a storage tank 18 for the inerting agent I. The vent line 14 opens at its end 20, designed as a Venturi scrubber, into a tank 22 which is an iodine and/or aerosol separating device and which has a filter element 24 in its upper region. An exhaust-gas line 26 is connected to the tank 22 on the outlet side via the filter element 24. The exhaust-gas line 26 opens into a stack 27, advantageously via a non-illustrated gas suction extraction device. Moreover, the vent line 14 is shunted in parallel with the tank 22 to the exhaust-gas line 26 via a bypass line 29. The bypass line 29 can be shut off with a valve 28. The activity holdup device 4 has a number of filter inserts 30, two of which are shown in FIG. 1. In this case, each filter insert 30 has adsorption material A on its surface. When it becomes necessary to inert the containment atmosphere CA, first the inerting agent I, which is present in liquid form in the storage tank 18 up to a level 32, is partially evaporated by means of a heating device 34. In this case, the inerting agent I may, for example, be nitrogen (N.sub.2), carbon dioxide (CO.sub.2), water or a mixture of these. The evaporated inerting agent I flows to the distributor 10 via the supply line 12 and also passes through the heat exchanger 16. The inerting agent I is superheated in the heat exchanger 16. The heat exchanger 16 may be an actively heated heating element or else a permanently heated heat energy reservoir (high temperature dry accumulator). To supply inerting agent I' superheated in this way into the interior of the containment 8, the passage of the supply line 12 to the line 2 is opened via the distributor 10, so that the superheated inerting agent I' is supplied to the activity holdup device 4 via the line 2. In the activity holdup device 4, the superheated inerting agent I' flows through the filter inserts 30 and then passes via the line 2 and the leadthrough 6 into the interior of the containment 8, where it contributes to inerting the containment atmosphere CA. After a quantity of superheated inerting agent I'--so much as to require a pressure reduction --has been supplied to the interior of the containment 8, the connection between the supply line 12 and the line 2 is closed and the connection between the vent line 14 and the line 2 is opened via the distributor 10. With the distributor 10 in this position, it becomes possible to vent the containment atmosphere CA, i.e., to bleed the containment atmosphere CA. In this case, containment atmosphere CA vented from the containment 8 flows through the activity holdup device 4 and its filter inserts 30. Radioactive material, such as, for example, inert gases, contained in the containment atmosphere CA is held up by adsorption on the filter inserts 30. Then, as indicated by the arrow, the filtered containment atmosphere CA' flows to the iodine and aerosol separator tank 22. Due to the interaction of the end 20 of the line 14 and the filter element 24, wherein the end is designed as a Venturi washer, iodine and/or aerosols are removed from the containment atmosphere CA'. The containment atmosphere CA", which has thus been further cleaned, is then released into the surroundings via the stack 27. After the pressure in the interior of the containment 8 has thus been sufficiently equalized, additional inerting agent I is supplied to the containment once again in a further step. For this purpose, the distributor 10 closes the connection of the line 2 to the vent line 14 and simultaneously opens the connection of the line 2 to the supply line 12. Superheated inerting agent I' thus flows once again into the interior of the containment 8 and, in this case, once more passes the activity holdup device 4 and its filter inserts 30. When the inerting agent I' flows through the filter inserts 30, it releases the radioactive material absorbed on these and conveys it back into the interior of the containment 8. Operating the apparatus 1 alternatingly and therefore discontinuously in this way reliably avoids the release of radioactive material even when the containment 8 is being vented. Referring now to FIG. 2, in the alternative apparatus 1' for inerting and venting the containment atmosphere CA, the supply line 12 for the inerting agent I and the vent line 14 for containment atmosphere CA are connected to the interior of the containment 8 through the leadthroughs 40 and 41, respectively. In this embodiment too, the supply line 12, into which the superheater heat exchanger 16 is inserted, is connected to the storage tank 18 for the inerting agent I. The vent line 14 is connected to the stack 27 via the tank 22, which is an iodine and aerosol separating vessel. In this exemplary embodiment, the supply line 12 and the vent line 14 are led via a joint reversible activity holdup device 4'. Here, the activity holdup device 4' is designed as a regenerating wheel. It comprises a filter element 52 which is mounted rotatably about an axis 50 and which also is provided with an adsorption agent A. The activity holdup device 4' can be operated continuously. Containment atmosphere CA vented from the containment 8 flows through part of the filter element 52, the part being located in the region of the vent line 14. Only this part of the filter element 52 is laden with radioactive material contained in the containment atmosphere CA. Similarly to the exemplary embodiment according to FIG. 1, the filtered containment atmosphere CA' subsequently flows to the tank 22, which is an iodine and/or aerosol separating device. Due to the rotation of the filter element 52 about the axis 50, that portion of the filter element 52 which is laden with radioactive material is removed from the region of the vent line 14 and is replaced by a portion of the filter element 52 which is not laden with radioactive material. Partial areas of the filter element 52 are sealed off relative to one another with activity-resistant and noncombustible sealing material. Due to the rotation of the filter element 52 about the axis 50, that portion of the filter element 52 which is laden with radioactive material passes into the region of the supply line 12 for the inerting agent I. There, inerting agent I' (superheated in the heat exchanger 16 and to be supplied to the containment 8) flows through that filter portion which is laden with radioactive material. As a result, previously adsorbed radioactive material is released from the filter element 52 and flushed back into the containment 8. Each portion of the filter element 52 is thus continuously laden with radioactive material and subsequently regenerated. The operations of charging and regenerating the filter element 52 thus take place in parallel and simultaneously in the activity holdup device 4', so that the latter can be operated continuously. This rules out a buildup of excess pressure in the containment 8 at all times, so that inerting the containment 8 in this way is particularly flexible and may also take place as a preventive measure (prophylactic) at any time. The inerting agent I provided in the exemplary embodiment according to FIG. 2 and stored in the storage tank 18 is water. This water I is evaporated completely or partially by means of the heating device 34. Steam D flowing to the activity holdup device 4' via the supply line 12 is superheated in the heat exchanger 16, so that particularly effective regeneration of that part of the filter element 52 through which the steam flows becomes possible. The steam D supplied condenses within the containment 8. This condensation, illustrated by the shading K, results in a pressure reduction or vacuum within the containment 8. It thereby becomes possible, particularly in combination with the stack 27, to maintain a vacuum permanently in the interior of the containment 8. By maintaining a vacuum of this kind, a discharge of radioactive material into the environment is reliably avoided, even in the case of a nontight containment 8 or in the event of leakage, since, as indicated by the arrows L in FIG. 2, leakages flow solely into the containment 8. A further supply line 54 for supplying a further inerting agent I.sub.2 may be connected to the activity holdup device 4' in parallel with the supply line 12. In this case, the inerting agent I.sub.2 may be nitrogen (N.sub.2) or carbon dioxide (CO.sub.2). This arrangement of this type makes it possible to regenerate the filter element 52 by means of a mixture of steam D and inert gas I.sub.2. Each filter element 30, 52 of the activity holdup device 4 or 4' may advantageously have activated charcoal and/or a molecular sieve as adsorption material A. In this case, the finely distributed adsorption material has an inner exchange surface of at least 1000 m.sup.2 /m.sup.3. For particularly effective regeneration of the filter elements 30 or 52, the temperature of the inerting agent I', I.sub.2 supplied can be regulated. Temperature regulation of this type may be carried out, for example, by regulating the heat exchanger 16 by means of the control device 15 or else, in a way not illustrated in any more detail, by dividing the gas stream carried in the supply line 12 into part streams, only one of which is led via the heat exchanger 16 and is subsequently admixed again with the other part streams. According to the exemplary embodiments shown in FIGS. 1 and 2, the activity holdup device 4, 4' is arranged outside the containment 8, but it may alternatively also be arranged within the containment 8. In order to ensure that the containment 8 is closed off reliably and a discharge of radioactive material into the surroundings is thus safely avoided, the supply line 12, following the lead through 6 or 40 through the containment 8, is provided with a selfclosing shutoff fitting 60, 60'. This is opened, for example, by means of the feed pressure of the inerting agent I', I.sub.2, said feed pressure acting counter to a spring force or weight. Thus, in the event of a failure or termination of the supply of inerting agent I', I.sub.2, the containment 8 is closed off relative to the surroundings on a "failsafe" principle. By means of an apparatus 1 or 1' of this type, it becomes possible at any time to inert the containment atmosphere CA and simultaneously vent the latter. Since an excess pressure buildup within the containment 8 is avoided, inerting in this way is possible at all times. Inerting may therefore take place after a "blowdown" or according to other criteria, such as, for example, "H.sub.2 concentration too high", "core filling level too low", or else as a preventive measure, since under no circumstances is there any fear that radioactive material will be released into the environment. Moreover, it is possible for the containment atmosphere CA to be inerted for a short time and even in the event of a failure of important systems (station blackout). |
description | This application claims priority to U.S. Provisional Patent Application No. 62/611,754, filed on Dec. 29, 2017 and entitled: AIR COOLING CHAMBER FOR CRDM DRIVE MOTOR, which is herein incorporated by reference in its entirety. This invention was made with Government support under Contract No. DE-NE0000633 awarded by the Department of Energy. The Government has certain rights in this invention. This disclosure generally relates to cooling in a containment vessel of a nuclear reactor module, and some embodiments relate to a nuclear reactor module with a cooling chamber for a drive motor of a CRDM (control rod drive mechanism). Convective heat transfer is the transfer of heat from one place to another by the movement of fluids (liquids or gasses). Convective heat transfer may include both forced convection (pumps to move liquid through hoses to carry away heat from a source, fans to drive the movement of air over fins or the like to carry away heat from the source, etc.) and natural convection (in which buoyancy forces that result from density variations drive the movement of the fluid). Some nuclear reactor modules include a reactor pressure vessel (RPV) housed inside a containment vessel (CNV). These nuclear reactor modules may include a reactor component cooling water (RCCW) system to support components inside and outside the reactor module. The RCCW system may include pumps and cooling lines outside of the CNV to dissipate heat from components outside the CNV. The RCCW system may also include cooling lines penetrating the CNV—to cool components within the CNV. Reduction of cooling lines inside the CNV (such as cooling lines of an RCCW system outside the CNV or any other cooling line inside the CNV) is desirable for a number of reasons. The environment inside the CNV may be depressurized (e.g., in vacuum, meaning less than atmospheric pressure) and high temperature (e.g., 600 degrees F.), and seals between these openings and the cooling lines provide leak paths for atmosphere to leak into the CNV resulting in outage time and loss of power generation. Also, each cooling line consumes valuable space inside the CNV—reducing cooling lines inside in the CNV may support reduction of the overall dimensions of the CNV. Cooling lines may also break, and diagnosing and/or repairing any cooling line, particularly those within the CNV, may be expensive. Some known nuclear reactor modules may include an RCCW system external to the CNV, including cooling lines extending into the CNV to dissipate heat from CRDM magnetic coils and/or CRDM drive motors. Some embodiments described herein eliminate some or all of such cooling lines by partitioning a section (e.g., an upper section) inside the CNV into a vacuum chamber for CRDM coils and a fluid-filled chamber for CRDM drive motors or other coils. The section may be partitioned using any fluid-tight partition. Any fluid-tight partition described herein may allow for differential movement between the CRDMs and the CNV. In some embodiments, an attachment section of the fluid-tight partition (to attach to the CRDM) may move relative to the CNV (e.g., with the CRDM housings). In some embodiments, the attachment section may be a rigid plate with an array of openings with each opening around one of the CRDM housings. The attachment section may be sealingly coupled to the CRDM drive by welding, o-rings around the CRDM housings, or the like, or combinations thereof. The fluid-tight partition may include an expansion section, such as a bellows (e.g., metal bellows), membranes, expansion joints, omega seals, or the like, or combinations thereof, to expand in response to movement of the CRDM housings towards the CNV, and contract in response to movement of the CRDM housings away from the CNV. The fluid-filled chamber may contain the drive motors and/or air (or some other fluid) that is thermally coupled to the drive motors. Natural and/or forced convective heat transfer may be used to remove heat from the drive motors using the air or other fluid in the fluid-filled chamber. When the CRDM housing is maneuvered using a fine motion control drive using an electric motor on top of the CRDM, heat generated by the energy supplied to the motor may be removed by an atmospheric air cooling environment. In some embodiments, the fluid-supply system may be an air-supply system, which may be substantially less expensive and reliable than an RCCW and may eliminate the need for hoses and piping inside containment. In embodiments with forced convection, a fluid-supply system may be placed on top of the nuclear reactor module or located in a common area remote from the CNV. In some embodiments, penetrations on top of the CNV (or some other location corresponding to the fluid-filled chamber) may provide supply air and remove air as necessary in a forced flow system or enable natural convection to cool the drive motors. These penetrations may be isolated from the vacuum chamber by the fluid-tight partition, and as such, may not require seals (in fact, in natural convection embodiments these penetrations may be unrestricted vents). As far as the vacuum chamber, by way of background, recent development in high temperature coil technology may allow placement of magnetic coils in a harsh environment similar to what is experienced inside some CNVs. These magnetic coils may be referred to as high temperature magnetic coils. High temperature magnetic coils may be located in the vacuum chamber and not energized to the levels that would require heat removal by water cooling. Water or other form of cooling to the coils may be eliminated. In other embodiments, any coils may be used in the vacuum chamber if their energy input is low and/or intermittent (e.g., if they are only energized intermittently—not long enough to increase their temperature beyond their operating range). In these embodiments, water cooling to the coils may be eliminated too. Some embodiments described herein may include a nuclear reactor module with a cooling chamber for a drive motor of a control rod drive mechanism. This nuclear reactor module may include a containment vessel for a reactor pressure vessel (RPV); a control rod drive mechanism (CRDM) located in the containment vessel, the CRDM including drive motors configured to move control rods into and out of a nuclear reactor core located in the RPV; and a partition extending across a portion of the containment vessel configured to retain the drive motors in a separate fluid-tight barrier region within the containment vessel. FIG. 1 illustrates a cross-sectional view of a containment vessel 2 with a cooling chamber 11 for a drive motor 17 of a CRDM (control rod drive mechanism) 15, according to various embodiments. The cooling chamber 11 may be separate from the chamber 10, which may be a vacuum chamber (e.g., depressurized to less than atmospheric pressure). The CRDM 15 may include high temperature magnetic coils 16 (or some other coils) that can operate in the high temperature environment of the chamber 10 (e.g., 600 F) without water cooling. The CRDM 15 and/or the shafts 3 inside housings, respectively, of the CRDM 15 may move relative to the containment vessel 2 and the RPV 1 to control movement of rods (not shown, each rod is connected to a corresponding one of the shafts 3) into and out of a nuclear reactor core (not shown) located in the RPV 1. This movement may be driven by the drive motor 17. During movement, the drive motor 17 may generate heat, and this heat may be removed by a fluid forced (e.g., by a fan, pump, or some other active component, not shown) or drawn in through the opening 19 (heated fluid may exit another opening). Unlike the openings for cooling lines in some containment vessels, the opening 19 is not exposed to a vacuum environment inside the nuclear reactor module. The fluid-tight barrier 13 isolates a portion of the containment vessel 2 from the vacuum environment, and the opening 19 is formed in this portion of the containment vessel 2. In this embodiment, the fluid-tight barrier 13 comprises an expansion section 14 sealingly coupled (e.g., welded) to an interior of the containment vessel. In this embodiment, the expansion section 14 is illustrated as a bellows, but in other embodiments the expansion section may comprises membranes, expansion joints, omega seals, a bellows, or the like, or combinations thereof. FIG. 2 illustrates a bottom view of the fluid-tight barrier 13. Referring now to FIG. 2, the fluid-tight barrier 13 may include an attachment section 251 connected to the other side of the expansion section 13. The attachment section 251 may include a rigid structure, such as a plate. An array of openings 252 may be formed in the attachment section for each of the CRDM housings (FIG. 1). The attachment section 251 and each CRDM housing may be sealingly coupled (e.g., welded and/or coupled using o-rings) where they contact in the openings 252. Referring again to FIG. 1, a portion of the fluid-tight barrier 13 (e.g., including the attachment section 251, FIG. 2) may move with the CRDM 15 relative to the containment vessel 2 and the RPV 1. The expansion section 14 may expand when the CRDM 15 moves toward the RPV 1, and compress when the CRDM 15 moves away from the RPV 1. FIG. 3 illustrates a cross-sectional view of another containment vessel 302 with a cooling chamber 311 for a drive motor (not shown) of a CRDM 315, according to various embodiments. The CRDM 315 may be similar in any respect to any CRDM described herein, such as the CRDM 15 of FIG. 1. The fluid-tight barrier 314 may be similar in any respect to any fluid-tight barrier described herein, such as the fluid-tight barrier 14 of FIG. 1. An attachment section 351 may be similar in any respect to any attachment section described herein, such as the attachment section 251 of FIG. 2. A portion of the drive motor may be below the attachment section 351 so long as the drive motor is thermally coupled to the fluid of the cooling chamber 311. For instance, the drive motor may be part of a drive motor assembly, and the attachment section 351 may be welded to the drive motor assembly. The drive motor assembly may include a fluid tight housing that exposes the drive motor to a fluid of the cooling chamber 311 to thermally couple the drive motor to the fluid and to isolate the fluid from the vacuum environment 310. In one example it may be possible and practical to weld the attachment section 351 to an end of the CRDM at the drive motor assembly. FIG. 3 also illustrates that an ingress hose 369 may be coupled to an ingress opening for the cooling chamber 311. An active component such as a fan may force air into the ingress hose 369. The heated air may exit through an egress hose 370 coupled to an egress opening for the cooling chamber 311 due to a positive air pressure created by the active component and/or natural convection (the egress opening may be located higher than the ingress opening to aid thermal transfer). In other embodiments, an active component may be installed at the egress hose 370 to create negative pressure in the cooling chamber 311, which may draw air into the ingress house 369. In contrast to water cooling lines that extend inside a containment vessel, hoses 360 and 370 may be entirely located outside the CNV 302. In some examples, hosing may be used for only one of ingress/egress, and the other opening may comprise a vent with no hosing. In yet other examples, all openings may be vents with no hoses. Also, there may be any number of egress/ingress openings/hoses/fans/vents, etc. Also, as mentioned, an active component may not be required if natural convection is adequate to remove heat. Also, in some examples, it may be possible and practical to install insulation along the fluid-tight barrier 314 (on one or both sides) to reduce heat transfer from the high temperature vacuum environment of the chamber 310 into the cooling chamber 311, to minimize the amount of heat to be removed through the egress opening. FIG. 4 illustrates a cross-sectional view of yet another containment vessel 402 with a cooling chamber 411 for a drive motor of a CRDM 415, according to various embodiments. The CRDM 415 may be similar in any respect to any CRDM described herein, such as the CRDM 15 of FIG. 1. The fluid-tight barrier 414 may be similar in any respect to any fluid-tight barrier described herein, such as the fluid-tight barrier 14 of FIG. 1. An attachment section 451 may be similar in any respect to any attachment section described herein, such as the attachment section 251 of FIG. 2. In this example, a heat exchanger 499 (schematically illustrated) may be coupled to (or formed) from the portion of the containment vessel 402 associated with the cooling chamber to remove heat generated by the drive motor of the CRDM 415 by thermal conduction, convective heat transfer, thermal radiation, or the like, or combinations thereof. The containment vessel 402 may or may not include openings (a heat exchanger component such as a heat pipe may be positioned in an opening). The heat exchanger 499 may include any components of any known heat exchangers and/or part of the containment vessel itself (the containment vessel wall may provide conductive heat transfer, for example). References have been made to accompanying drawings, which form a part of the description and in which are shown, by way of illustration, specific implementations. Although these disclosed implementations are described in sufficient detail to enable one skilled in the art to practice the implementations, it is to be understood that these examples are not limiting, such that other implementations may be used and changes may be made to the disclosed implementations without departing from their spirit and scope. Example 1 is an apparatus comprising a containment vessel for a reactor pressure vessel (RPV); a control rod drive mechanism (CRDM) located in the containment vessel, the CRDM including drive motors configured to move control rods into and out of a nuclear reactor core located in the RPV; and a partition extending across a portion of the containment vessel configured to retain the drive motors in a separate fluid-tight barrier region within the containment vessel. Example 2 may include the subject matter of example 1 and/or any other example herein, wherein the partition comprises: an attachment section having a side sealingly coupled to the CRDM; and an expansion section coupled to another side of the attachment section, the expansion section to expand to permit the attachment section to move with the CRDM relative to the containment vessel. Example 3 may include the subject matter of any of example 1 and/or any other example herein, wherein the expansion section comprises a bellows, a membrane, an expansion joint, or an omega seal. Example 4 may include the subject matter of any of examples 1-3 and/or any other example herein, wherein the attachment section comprises a plate. Example 5 may include the subject matter of any of examples 1-4 and/or any other example herein, wherein the expansion section is sealingly coupled to the portion of the containment vessel. Example 6 may include the subject matter of any of examples 1-5 and/or any other example herein, wherein separate fluid-tight barrier region within the containment vessel comprises a first region, and wherein a second region of the containment vessel comprises a vacuum environment. Example 7 may include the subject matter of any of examples 1-6 and/or any other example herein, wherein the partition isolates the portion of the containment vessel from the vacuum environment, and wherein the portion of the containment vessel comprises one or more openings to exchange air outside the containment vessel with air inside the separate fluid-tight barrier region. Example 8 may include the subject matter of any of examples 1-6 and/or any other example herein, further comprising an active device located outside the containment vessel to force the air outside the containment vessel through an ingress one of the one or more openings into the separate fluid-tight barrier region or force the air inside the separate fluid-tight barrier region through an egress one of the one or more openings to move heat generated by the drive motors out of the containment vessel. Example 9 may include the subject matter of any of examples 1-8 and/or any other example herein, wherein the separate fluid-tight barrier region comprises air at a higher density than any air located in a vacuum region of the containment vessel. Example 10 may include the subject matter of any of examples 1-9 and/or any other example herein, further comprising a heat exchanger located on the portion of the containment vessel to remove heat generated by the drive motors from the containment vessel. Example 11 is an apparatus, comprising: a containment vessel including a first section to house a reactor pressure vessel of a nuclear reactor module and a second different section above the first section; a control rod drive mechanism (CRDM) located in the second section, the CRDM to control movement of the CRDM relative to the containment vessel; a fluid-tight barrier around the CRDM, the fluid-tight barrier and the CRDM partitioning the second section into a vacuum chamber and a fluid-filled chamber, wherein at least a portion of the CRDM is positioned in the vacuum chamber; wherein one or more CRDM drive motors of the CRDM are thermally coupled to a fluid of the fluid-filled chamber. Example 12 may include the subject matter of any of example 11 and/or any other example herein, wherein the fluid-tight barrier comprises an attachment section sealingly coupled to the CRDM and an expansion section to expand to permit the attachment section to move with the CRDM relative to the containment vessel. Example 13 may include the subject matter of any of examples 11-12 and/or any other example herein, wherein the attachment section comprises a plate with an opening for each CRDM housing of the CRDM. Example 14 may include the subject matter of any of examples 11-13 and/or any other example herein, wherein the expansion section comprises a bellows, a membrane, an expansion joint, or an omega seal. Example 15 may include the subject matter of any of examples 11-14 and/or any other example herein, wherein the attachment section is welded to the CRDM. Example 16 may include the subject matter of any of examples 11-15 and/or any other example herein, wherein the at least the portion of the CRDM comprises one or more CRDM magnetic coils. Example 17 may include the subject matter of any of examples 11-16 and/or any other example herein, wherein the at least the portion of the CRDM comprises a first section of the CRDM and a second different section of the CRDM is positioned in the fluid-filled chamber. Example 18 may include the subject matter of any of examples 11-17 and/or any other example herein, wherein one region of the fluid-tight barrier is sealingly coupled to an interior of the containment vession and another different region of the fluid-tight barrier is sealingly coupled to the CRDM. Example 19 may include the subject matter of any of examples 11-18 and/or any other example herein, wherein the fluid comprises air. Example 20 may include the subject matter of any of examples 11-19 and/or any other example herein, further comprising plural fluid exchange openings in the containment vessel to form fluid paths in and out of the fluid-filled chamber. Having described and illustrated the principles of a preferred embodiment, it should be apparent that the embodiments may be modified in arrangement and detail without departing from such principles. Claim is made to all modifications and variation coming within the spirit and scope of the following claims. |
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summary | ||
description | This patent application incorporates by reference U.S. provisional patent application 62/002,763, filed May 23, 2014, and U.S. provisional patent application 62/046,453, filed Sep. 5, 2014, which are incorporated by reference along with all other references cited in this application. The present invention relates generally to providing a portable ultraviolet (UV) light source for curing UV-curable gel nail polish. More particularly, the present invention relates to a portable UV nail lamp with a surface-mounted light emitting diode (SMD LED) light source. The present invention also relates to a UV nail lamp with a light emitting diode (LED) light source and a platform for a user's hand. UV nail lamps are available for the salon and home to cure UV-curable nail polish. These nail lamps typically have UV fluorescent tubes or bulbs that use alternating current (AC) power. So, these nail lamps have an AC cord that needs to be plugged into the wall, which restricts their placement, since they need to be close to a wall socket. This can be problematic. In a salon, for example, this can restrict the number of lamps in use, the location of nail lamp stations, and thus, the number of customers that can use the lamps at a given time. The tubes or bulbs of these nail lamps consume rather significant amounts of power and generate heat, which makes these nail lamps typically large and bulky to accommodate the bulb size and to allow for heat dissipation. This makes these nail lamps somewhat difficult to move, and certainly very difficult to travel with and use in a location without a wall socket, such as while on an airplane. Further, the light from the bulbs of these lamps tends be uneven, so a person's nails are exposed to difference intensities of light output, which causes the nails to dry at different times or to cure unevenly. Further, traditional nail lamps use light bulbs that tend to produce uneven light, so a person's nails are exposed to difference intensities of light output, which causes the nails to dry at different times or to cure unevenly. These bulbs also tend to be bulky which causes the nail lamps to be large and cumbersome. Conventional bulbs can also consume much electrical energy while operating. These lamps often have a flat platform on an inside of the lamp for a user to place their hand during drying. With long drying times, the user's hand can become uncomfortable or cramp up with the fingers in a strained, stretched out position within the lamp. There is a risk that the nails can smudge before setting as the user's nails brush up against other fingers or inside the lamp. As can be appreciated, an improved nail lamp is needed. What is also needed is a method and an apparatus which can accommodate a user's five fingers in a comfortable and ergonomic resting position within a nail lamp. What is also desired is an efficient way to evenly cure UV-curable nail polish on each of the user's nails. A nail lamp for curing UV-curable nail gel is powered by direct current (DC) and can be battery operated. The nail lamp uses surface-mounted light emitting diodes (SMD LEDs) which are relatively lower power. The nail lamp is easily transportable and can be used even when a wall socket is unavailable, such as while traveling on an airplane or in a car. The nail lamp has a cavity or treatment chamber that can accept a user's five fingers. So, the nail lamp can evenly cure nail polish on up to five fingers at once. A compact portable LED nail curing lamp has surface-mounted light emitting diode (SMD LED) lights. The lamp provides fast and consistent results producing high gloss finish and even curing of nail polish (e.g., UV-curable gel polish). The nail lamp has a micro-USB port, which can be used to power the lamp using a wall adapter, car charger, laptop USB port, or mobile power bank for ultimate portability. In an implementation, a system includes a compact LED nail curing lamp and a mobile power battery pack. The system also includes a cable to connect the nail lamp and the mobile power battery pack. The battery pack provides portable power to the nail lamp so that the nail lamp can be used portably, such as during travel or on an airplane when a wall outlet is unavailable. A compact LED nail curing lamp has a sleek design with advanced technology, highly efficient surface-mounted light emitting diode (SMD LED) lights. The lamp provides excellent results producing high gloss finish and even curing of nail polish (e.g., UV-curable gel polish). A specific implementation of a compact LED nail curing lamp is the SMD LED Lamp S2 product by LeChat Nail Care Products of Hercules, Calif. The compact LED nail curing lamp has a micro USB port, which is convenient to use. The user can power this SMD LED lamp (e.g., LeChat's LED Lamp S2 product) using a wall adapter (included), car charger (optional), laptop USB port, or mobile power bank for ultimate portability. In an implementation, a mobile power bank battery that can be used with the SMD LED Lamp S2 product is the LeChat Mobile Power™ battery pack by LeChat Nail Care Products. This product is approved by the Underwriters Laboratories. The packaging of the product can include the certification “UL Approved.” The product is also compliant with U.S. and international standards of the Restriction of Hazardous Substances Directive (RoHS) for environmental friendly products. In an implementation, a system includes a compact LED nail curing lamp (e.g., LeChat S2 product) and a mobile power battery pack (e.g., LeChat Mobile Power product). The system also includes a cable to connect the nail lamp and the mobile power battery pack. In an implementation, the nail lamp has a micro-B USB connector input and the mobile power battery pack has a type A USB receptacle, and the cable connects these together. The battery pack provides portable power to the nail lamp so that the nail lamp can be used portably, such as during travel or on an airplane when a wall outlet is unavailable. The lamp has a large, illuminated single-button that turns the lamp on for a preset cure time of 30 seconds for efficient, rapid LED/UV gel curing. The compact design saves space and allows for portability that is convenient for travel and pedicure applications. The lamp is lightweight and designed for carrying from place to place. The nail lamp includes professional durable materials that are long lasting and reliable. In an implementation, the nail lamp is a 6-watt LED lamp that includes forty-two SMD LED lights that provide evenly distributed light that allows for an efficient cure in about 30 seconds. In an implementation, a system includes: a upper housing having a button and a power input; and a lower housing, connected to the upper housing, the cavity or treatment chamber including openings through which surface-mounted light emitting diodes can emit light through. The cavity is sufficiently wide (e.g., about 4.25 inches or 10.6 centimeters) to accommodate five fingers of a human hand placed on a flat surface. In an enclosure formed between the upper and lower, there is circuitry. The circuitry includes at least one printed circuit board with the surface-mounted light emitting diodes; a button; a multiplexer, connected to the power input; a control circuit, connected to button and the multiplexer; a timer, connected to the control circuit and the multiplexer; a recharging circuit, connected and the multiplexer. The system includes a rechargeable battery comprising a battery output coupled to the multiplexer. The recharging circuit is connected to the rechargeable battery, so it can be recharged from, for example a wall outlet, that is connected to the power input. The multiplexer switches between the power input and the rechargeable battery to supply power circuitry. The housing can include a USB power output, which can be used to power or charge other devices. The power input can be a micro USB power input, which is readily available. A nail lamp includes a housing including a base and an outer cover. On a front side of the housing, there is an opening to a cavity within the housing. Inside the housing are inner surfaces of the housing including a platform, an inner side wall, and an inner roof of the housing. The opening is shaped and sized to allow a user's hand or foot to pass through the opening into the space within the housing. A finger plate is positioned on an inside of a housing of a nail lamp. The finger plate includes five side by side depressions that are adapted to support a user's fingers when the user places a hand inside the housing on the plate. In an implementation, the finger plate is removable from the housing. Different finger plates (or foot plates) can be used for users with different size hands or feet. An arrangement of light sources is positioned on sidewalls and inner roof of an inside of a housing. The light sources can be LEDs using surface mount technology (SMT), or surface mount devices (SMD) LEDs. In an implementation, a SMD LED can produce UV light in a range of about 340 nanometers to about 410 nanometers. Other objects, features, and advantages of the present invention will become apparent upon consideration of the following detailed description and the accompanying drawings, in which like reference designations represent like features throughout the figures. FIGS. 1-8 show views of a nail lamp 100. FIG. 1 shows a perspective view, FIG. 2 shows a top view, FIG. 3 shows a front side view, FIG. 4 shows an upside down view, FIG. 5 shows a right side view, FIG. 6 shows a back side view, FIG. 7 shows an inside view, and FIG. 8 shows the nail lamp as part of a kit 800. The nail lamp device has an exterior surface 102 and at one side, an opening 104 through which a user can place their hand into an interior space 106 of the nail lamp. There is a control button on the exterior that is used to turn on an interior lighting source 108 of the device, which exposes the interior space to light from the interior lighting source. As an example, a user can insert their fingers into the interior space, turn on the cure interior lighting source, and cure their UV nail polish or UV nail gel coated nails with the interior light. In an implementation, there is also an exterior lighting source (e.g., an LED) of the device, which also turns on in response to the control button and is on when the interior lighting source is on. Light from the exterior lighting source is visible through a translucent material (e.g., translucent plastic) of the control button. When the interior lighting source is off, the light from the exterior lighting source will also be off. The exterior lighting source is used as an indicator that the device is on—that the interior lighting source is on. In an implementation, the interior lighting source emits light of a different wavelength from the exterior lighting source. The interior lighting source can emit UV light (wavelengths ranging approximately from 100 nanometers to 400 nanometers) to cure UV-curable gel polish. And the exterior lighting source emits wavelengths of light within the visible light spectrum (wavelengths ranging approximately from 390 nanometers to 700 nanometers). In specific implementations, the exterior lighting source emits red, green, blue, or any combination of red, green, or blue colors. The red colors include wavelengths ranging approximately from 620-740 nanometers. The green colors include wavelengths ranging approximately from 495-570 nanometers. The blue colors include wavelengths ranging approximately from 450-495 nanometers. More specifically, the nail lamp includes a housing. The housing includes an outer cover (also be referred to as an exterior surface) and inner walls. In an implementation the outer cover is made a plastic material that has a glossy sheen finish (e.g., metallic finish). On a side of the housing, there is an opening to a space (or cavity or interior space or treatment chamber) within the housing. The space within the housing is defined by inner walls of the housing. The inner walls can be made of a reflective material. This material can direct emitted light from SMD LEDs into the cavity toward the user's nails. In an implementation, the interior of the lamp includes six inner walls. One of the walls forms a ceiling of the cavity. The other walls are angled with respect to this wall. In another implementation, shown in FIG. 4, the interior of the lamp includes seven inner walls, 110, 112, 114, 116, 118, 120, and 122. In an implementation, the opening is shaped and sized to allow a user's hand to pass through the opening into the cavity. In another implementation, the opening is adapted to allow a foot to pass through the opening. In another implementation, the nail lamp is adapted to be used for both a hand and foot. FIG. 6 shows a specific implementation of a nail lamp that includes a port 124 for a micro-USB connector cable. A power source can be coupled to the port to provide the nail lamp with operating power. In other implementations, the port can be a USB port, or plug, or other types of ports for electrical power transfer. On a bottom of the housing, there are grip members that prevent the housing from sliding on a work surface. The grip member is one or more rubber pads which provide friction against the surface. The grip members can help stabilize the nail lamp during curing to prevent nudging the nails during use or on uneven or unlevel surfaces (e.g., table on a train or airplane). FIG. 8 shows a specific implementation of a nail lamp that is part of a kit. The kit includes a packaging (e.g., a box) that includes the nail lamp 100, a power adaptor 128, and a USB/micro-USB cable 130. Below is a table of operational modes of the SMD LED lamp. ModeOperational Mode1. No power to power inputUV light is not operational2. Power to power inputPower UV light components andoperational3. Press button when UV light offUV light turns on and turns offautomatically after 30 seconds(or other preset time)4. Press button while UV light onUV light immediately turns off FIG. 9 shows a block diagram of a cross-section of a nail lamp 900. There are five inner walls of the cavity that are visible. There is a first wall 902 that forms a ceiling of the cavity. There are two walls 904 and 906 next to the right and left of the first wall that are angled with respect to the first wall. The first, second, and third walls have SMD LEDs 907 that are attached to printed circuit boards arranged between these inner walls and the outer cover. The cavity also includes a fourth wall 908 adjacent the second wall and a fifth wall 910 adjacent the third wall. These walls have a reflective material 912 (e.g., iron, steel, aluminum, aluminum alloy, other metal or metal alloy, or other sheet metal) to direct 913 light into the cavity, and do not include SMD LEDs. A button 914 is coupled to an exterior 916 of the nail lamp. FIG. 10 shows a block diagram of a specific implementation of a first printed circuit board 1000 (PCB1). A power input 1002 (e.g., a universal serial bus (or USB) power connector input) provides power to a timer 1004, a control circuit 1006, and a LED driver 1008 of PCB1. A button 1010 is connected to the control circuit that is connected to the timer. The button can activate the control circuit that controls the timer which activates the LED driver to activate one or more SMD LEDs 1012 of PCB1. The LED driver can also control an LED 1014 that connects to the button. For example, the LED will turn on behind the button to cause the button to light up. FIG. 11 shows a block diagram of a cross section of a double-sided printed circuit board PCB1 1100 with SMD LED lights 1102 and 1104 attached to opposite sides of PCB1. There are two SMD LEDs 1102 on one side of PCB1 that emit light in a first direction away from PCB1 toward a button 1106 of the nail lamp (e.g., a back-lit control button). On an opposite side of PCB1, there is a group of SMD LEDs 1104 that emit light in a second direction away from PCB1 into a cavity of the lamp housing. FIGS. 12A-12B shows a comparison between a standard LED 1202 and a SMD LED 1204. Light from a standard LED is emitted at a smaller beam angle (angle A) compared to the SMD LED which has a greater beam angle (angle B) and beam spread. At a given distance away from a surface, the SMD LED and standard LED will each emit light in the shape of a cone. The SMD LED has a greater beam spread and will emit a greater area of illumination than the standard LED. So, a base of the cone of light (e.g., circle) for the SMD LED will have a greater area (e.g., greater diameter, B is greater than A) than that of a standard LED. Thus, fewer SMD LEDs are needed to light an area, allowing for less power used and greater energy savings. FIG. 13 shows a block diagram of a specific implementation of a nail lamp 1300 with four internal printed circuit boards. PCB1 1302 is connected to a second printed circuit board PCB2 1304 and a third printed circuit board PCB3 1306. PCB2 and PCB3 each includes at least one SMD LED light. PCB1 is also connected to a fourth printed circuit board PCB4 1308, which includes a USB connector input 1310. PCBs 1-3 provide the SMD LEDs that light the UV light cavity of the nail lamp housing. The cavity has a top horizontal section (light provided by PCB1) and two angled sections (light provided by PCBs 2 and 3) relative to the top horizontal section. And a micro USB connector (provided by PCB4) is positioned at a back of the nail lamp housing. In a specific implementation, PCBs 1-3 provide 42 LEDs, of which 24 are on PCB1, 9 are on PCB2, and 9 are on PCB3. In a specific implementation, a compact LED nail curing lamp has a sleek design with advanced technology, highly efficient surface-mounted light emitting diode (SMD LED) lights. The lamp provides excellent results producing high gloss finish and even curing of nail polish (e.g., UV-curable gel polish). A specific implementation of a compact LED nail curing lamp is the SMD LED Lamp S2 product by LeChat Nail Care Products of Hercules, Calif. The compact LED nail curing lamp has a micro USB port, which is convenient to use. The user can power this SMD LED lamp (e.g., LeChat's LED Lamp S2 product) using a wall adapter (included), car charger (optional), laptop USB port, or mobile power bank for ultimate portability. In an implementation, a mobile power bank battery that can be used with the SMD LED Lamp S2 product is the LeChat Mobile Power™ battery pack by LeChat Nail Care Products. This product is approved by the Underwriters Laboratories. The packaging of the product can include the certification “UL Approved.” The product is also compliant with U.S. and international standards of the Restriction of Hazardous Substances Directive (RoHS) for environmental friendly products. In a specific implementation, the lamp has a large, illuminated single-button that turns the lamp on for a preset cure time of 30 seconds for efficient, rapid LED/UV gel curing. The compact design saves space and allows for portability that is convenient for travel and pedicure applications. The lamp is lightweight and designed for carrying from place to place. The nail lamp includes professional durable materials that are long lasting and reliable. In a specific implementation, the nail lamp is a 6-Watt LED lamp that includes forty-two SMD LED lights that provide evenly distributed light that allows for an efficient cure in about 30 seconds. An SMD LED is mounted and soldered into a circuit board. Compared to a standard LED, an SMD LED is small in size since it has no leads or surrounding packaging that a standard LED has. A SMD LED does not have the standard LED epoxy enclosure, and thus, SMD LED lights emit a much wider viewing angle instead of the focused, narrow light of the standard LED. SMD LEDs provide advantages over standard LEDs. The SMD LED has lower voltage and current requirements which allows it to give off very little heat. SMD LEDs emit a higher level of brightness while consuming less power than standard LEDs. With standard LEDs, the UV light produced to cure UV gels over time breaks down the epoxy surrounding the standard LED causing the epoxy to crack. Once cracked, the standard LED no longer flows evenly, which disrupts the transmission of light, resulting in an uneven cure. In contrast, SMD LEDs have no epoxy that surrounds it, and thus, will not crack. The resulting emission of light will be even throughout the lifetime of the light. Further, standard LEDs use a higher voltage and therefore, produce more heat. The heat produced by the higher voltage LED lights can shorten the life of the standard LED, which causes them to go out faster compared to SMD LEDs. In a specific implementation, the SMD LED Lamp S2 product is a nail lamp having a 6-Watt LED lamp with an output voltage of 5 volts and 1.2 amps. The lamp includes 42 SMD LED lights. A width of the lamp is about 103.5 millimeters. A length of the lamp is about 146.5 millimeters. A height of the lamp is about 56 millimeters. In an implementation, the nail lamp product is part of a kit which includes a universal AC adapter. The adapter has an input power of about 100 volts to about 200 volts at 50 or 60 hertz. The adaptor has an output power of about 12 volts at 1.2 amps. The kit also includes a user guide or manual which includes operating instructions, safety warranty, product specifications, a certificate of warranty, and a warranty registration card. To use the SMD LED Lamp S2 product, a user can follow the following instructions (which are included on the user manual): 1. Plug the power adaptor into the back of the SMD LED lamp and then plug the other end into a wall outlet, a car outlet, a computer, or a mobile power bank. 2. To turn the SMD LED lamp “on,” press the power button that is located on top of the lamp to the “on” position, where the LED light of the button lights up. The lamp will automatically shut off after 30 seconds. 3. The SMD LED lamp can be used with both fingernails and toenails. For toenails, the user can place the lamp over toes and perform steps 1 and 2 above. The user should follow the following safety precautions when using the SMD LED lamp product. These precautions are included on the user guide as part of the kit. 1. Never look directly into the LED/UV lights when machine is ON. 2. Do not overexpose the nails or skin under light. 3. Do not use the LED light in or around water. 4. Unplug the LED light when not being used. 5. Certain cosmetics or prescriptive lotions can cause sensitivity to LED light. Do not use lamp if using any. 6. Do not pull the cord to unplug. Instead, grab plug firmly and pull to unplug. 7. Do not use any corrosive sanitizer, solvents, thinners, or scrubbing to clean the machine. 8. Do not stack anything on top of the LED Lamp. 9. Do not disassemble the LED Lamp. This will void the Warranty. 10. Do not try to repair the machine. Please contact the distributor for service. 11. The plastic bag in packaging is a choking hazard. Do not place over head. Keep away from children and pets. 12. The electric power system is labeled on the box. Please pay attention to the voltage and frequency. FIG. 14 shows a block diagram of a specific implementation of a nail lamp that is adapted to be used with a rechargeable battery pack 1402 that is external 1404 to the housing 1406 of the nail lamp. The rechargeable battery is a unit that is separate from the nail lamp. Circuitry to recharge this rechargeable battery pack is contained within (or internal 1408 to) a housing of the rechargeable battery pack. There battery pack (or the nail lamp) may have a battery gauge or charge level indicator that indicates a charge level remaining in the battery. For example, the battery gauge can indicate there 75 percent charge remaining in the battery pack. For example, in an implementation, the display of the nail lamp can display the battery charge level of the battery pack (such as by the user pressing a battery charge level button). For example, the rechargeable battery is a portable power pack with a USB plug output (e.g., type A USB receptacle). The nail lamp has a USB power connector 1410 (e.g., micro-B USB receptacle) that can connect to the rechargeable battery using a cable. The micro-B USB receptacle of the nail lamp is connected to the type A USB receptacle of the rechargeable battery via a micro USB cable. Then, the battery pack supplies power to the nail lamp (which consumes 6 watts maximum). In an implementation, the nail lamp consumes 6 watts or less of power. Through the USB, the power adapter or batter can provide about 5 volts and 1.2 amps. In other implementations, the nail lamp consumes 5 watts or less of power (e.g., 5 volts and 1 amp), 4.5 watts or less (e.g., 5 volts and 900 milliamps), or 2.5 watts or less of power (500 milliamps). In another implementation, the nail lamp consumer more than 6 watts, such as 10 watts (e.g., 5.1 volts and 2.1 amps) or 12 watts (5.1 volts and 2.4 amps). With more power, the cavity of the nail lamp can be made larger (allow for more comfort or larger hands), or there can be more LEDs (for more even light coverage), or higher intensity LEDs (possibly for better nail curing), or any combination of these. Thus the nail lamp and rechargeable battery are a nail lamp system that allow for cordless (e.g., not connected to a wall outlet) and portable use. Users and customers need not rely on being within proximal distance to a wall outlet. In a salon, this can restrict the number of lamps in use, the location of nail lamp stations, and thus, the number of customers that can use the lamps at a given time. With a portable rechargeable nail lamp, salon customers can dry their nails anywhere in the salon, which allows for more customers that can be serviced at a given time, and reduced wait times for customers. Further, a portable rechargeable nail lamp is convenient to use during travel (e.g., on a train or airplane), and in places where there is limited or no access to wall outlets. Users can also save time by drying their nails while doing other tasks that would otherwise had to have been done at other times. For example, while working on a laptop or making phone calls at work, a person can concurrently cure their nails while the nail lamp is running on batteries or connected to their laptop. Although this application specifically describes the nail lamp as having a micro-B USB receptacle and the battery pack as having a type A USB receptacle, one having ordinary skill in the art understands that other connector types can be used to provide power. For example, some other connectors may be used such as mini-USB connector (e.g., USB mini-B), mini-A, micro-AB, or Apple's lightning connector. In a specific implementation, a portable external battery pack is the LeChat Mobile Power™. The Mobile Power pack product includes a battery housing having a USB output port, a micro USB input port, an LED power indicator, a power or flashlight button, and an LED light. The Mobile Pack product also includes a cable for connecting the battery housing with a nail lamp (e.g., the SMD LED Lamp S2 product). The cable includes a USB cable, a micro USB connector on one end of the cable, and a USB connector on an opposite end of the cable. To charge the Mobile Power product, a user can connect the micro USB connector of the cable to the micro USB input port of the external battery housing, and the other USB connector end of the cable to a USB port of a power source including a wall adapter (to a wall outlet), a laptop USB port, a desktop USB port, or a DC 5-volt USB charger. The LED power indicator of the battery pack will flicker to indicate that the external battery has started charging. When all LED power indicator lights are lit, this indicates that the battery is fully charged. In an implementation, there are four battery indicator lights arranged in a row on an external surface of the battery pack. When the Mobile Power battery pack is fully charged and ready to be used to power an electronic device, the user should first check whether the charging voltage of the digital or electronic device is matched with an output voltage (DC 5 volts) of the external battery. The user can connect the USB connector of the cable to the USB port of the battery pack, and the other micro USB connector end of the cable to a micro USB port of an electronic device such as the SMD LED nail lamp. The can be used as a general mobile power pack, and can be used to power other electronic devices such as a smart phone, tablet device, or any electronic device with a DC 5-volt USB input. A number of the battery LED power indicator lights will light according to the remaining charge capacity of the battery pack. In a specific implementation, there are four indicator lights (L1-L4) in a row with L1 on a left end, L2 to the right of L1, L3 to the right of L2, and L4 to the right of L3, and on the right end. When L1 is flashing, this indicates that there is about 0 to about 25 percent charge capacity level in the battery. When L1 and L2 are flashing, this indicates that there is about 25 to about 50 percent charge capacity level in the battery. When L1, L2, and L3 are flashing, this indicates that there is about 50 to about 75 percent charge capacity level in the battery. And when L1, L2, L3, and L4 are flashing, this indicates that there is about 75 to about 100 percent charge capacity level in the battery. When the capacity remaining in the battery is less than about 5 percent, the first light (L1) will blink to remind the user to recharge the external battery. In a specific implementation, the external battery includes a flashlight button for a flashlight function. To activate the flashlight option, the user can double click the flashlight (or power) button on the battery. Brightness of the light will cycle between 10 percent, 50 percent, and 100 percent brightness. The flashlight should not be turned on under hot temperature environments for long periods of time. In a specific implementation, when the power button is pressed, the LED indicator lights will turn on. These lights will automatically turn off in about 10 seconds for power saving. When needing to charge or power digital or electronic products, the user can simply plug the cable into the external battery device, and it will start charging when it detects the load. The user should follow the following safety precautions when using the Mobile Power product. These instructions are included in a kit containing the Mobile Power product. 1. Charge fully before using the mobile power device. 2. Do not place or use mobile device at high temperature or in humid environment. Do not expose to excessive sunlight. (Operating temperature range: charging: 0 degrees Celsius to 45 degrees Celsius; discharging: −10 degrees Celsius to about 60 degrees Celsius; and storage environment: about −20 degrees Celsius to about 60 degrees Celsius). 3. The user should not throw the mobile power device in fire or water so as to avoid fire, explosion, or both. 4. Keep the mobile power device out of reach of children. 5. Do not disassemble the device arbitrarily, since in some of the products, there are no removable or maintainable parts that are installed in the product. 6. Do not vigorously shake, hit or impact the mobile power device. 7. If the mobile power device has exposed liquid or other abnormalities, discontinue use, and contact customer service. 8. If the mobile power device has liquid leakage and splashes into the user's eyes, do not rub the eyes, wash with clean water immediately, and go to the hospital for medical treatment. 9. It is normal for the temperature of the mobile power device to rise during use; do not operate in a confined environment. 10. The transmission lines and connectors of the mobile power device must be provided by the original manufacturer. The use of transmission lines or connectors of non-original manufacturer may result in severe or fatal injuries and property losses. 11. Do not cover or block the mobile power device with paper or other objects, to avoid blocking the heat dissipation and cold cutting. 12. Do not use the mobile power device if nobody is watching it in the car or anywhere. 13. Before using mobile power device, check its voltage demand. 14. If the mobile power device is not used for a long period of time, please charge or discharge it once every three months to ensure service life. 15. Remove power supply and power cord when the mobile power device is not in use. 16. Fully charge the mobile power device after the mobile power device is fully discharged. FIG. 15 shows a block diagram of a specific implementation of a nail lamp 1500 having a PCB5 1502 that can receive power from a USB power connector 1504 (e.g., micro-B USB receptacle) or rechargeable battery pack 1506. Unlike the FIG. 14 system, the rechargeable battery pack is specifically adapted to connect directly to the nail lamp circuitry (powering the nail lamp) without using the USB power connector. Specifically, power is not provided from the battery pack through the USB power connector, but rather directly from the battery. Further, the rechargeable battery pack can integrate with the housing of the nail lamp. In an example, the rechargeable battery pack snaps into place into a bottom of the nail lamp via a latching mechanism. And the rechargeable battery pack can be unlatched to be removed and replaced with a new pack, which may be desirable when the pack is spent or no longer holding charge (e.g., at the end of life of the pack). In an implementation, compared to the FIG. 14 system, circuitry to recharge this rechargeable battery pack is contained within a housing of the nail lamp (e.g., PCB5 of the nail lamp). Referring to FIG. 16, PCB5 is similar to PCB1 as described previously, but includes a recharging circuit 1602 and other circuitry to multiplex 1604 (mux), switch 1606, or other switching mechanism to switch between taking power from the USB power connector or the rechargeable battery pack. Power from the USB power connector (such as connected to a wall adapter or other power source) can be used to power the nail lamp and also recharge (via the recharging circuit) the rechargeable battery too. FIG. 17 shows an implementation where the nail lamp of FIG. 16 includes one or more USB power output connectors 1701. These connectors can be used to charge a user's or customer's device, such as a phone or tablet. The user or customer will connect their device (e.g., phone) via a cable to one power output connectors. The device will be charged from the power from the USB power connector input 1702 or the battery 1703 through a mux 1704 or switch. Typically when the USB power input is connected to power, this power is used to charge the user's device (and also the rechargeable battery pack of the nail lamp). When the USB power input is not connected to power, the user's device is charged by the nail lamp battery. FIG. 18 shows an example of a rechargeable battery pack 1802 that can be connected 1803 to the housing of nail lamp 1804. In this implementation, the battery is contained within a base plate 1806 of the nail lamp. When the nail lamp is used, the user or customer places their fingers (that will be exposed to the UV light) onto the battery pack base plate. The battery pack base plate snaps or latches into place in the housing of the nail lamp. FIG. 19 shows an outline of a plan view of the battery pack base plate. More specifically, referring to FIG. 18, the rechargeable battery pack connects to the nail lamp at one or more connection points via connectors. For example, the nail lamp has a connector for connecting to the external rechargeable battery pack which the nail lamp is designed for. In a specific implementation, the nail lamp has a female connector while the external rechargeable battery pack has a corresponding male connector that fits into the nail lamp's connector. In another specific implementation, the nail lamp includes a male connector that fits into the external rechargeable battery pack's female connector. In other implementations, however, the nail lamp's connector can have any number or combination of pins and shapes in order to interface with the external rechargeable battery pack that the nail lamp is designed for. In a specific implementation, the nail lamp can include a fastening member that fastens to the external rechargeable battery pack to ensure a tight fit. As an example, the nail lamp can include a latch to secure the lamp to the battery. In another specific implementation, when the external rechargeable battery pack is connected to the nail lamp, the nail lamp looks for an authentication or handshaking signal (e.g., sending of an authentication code). If the lamp does not receive the proper authentication, the lamp may display a signal (e.g., flashing lights) that the battery is not an authorized peripheral for the lamp or the lamp can simply not allow the lamp circuitry to interface with the battery (e.g., not allow charging). An authentication circuit can be included in the circuitry of the lamp to provide proper authentication to the nail lamp. FIG. 19 shows a specific implementation an outline of a plan view of the battery pack base plate 1806 that is designed for a nail lamp. In an implementation, the nail lamp is the SMD LED Lamp S2 product by LeChat Nail Care Products. The shape of the external rechargeable battery pack corresponds to the shape of a base of the nail lamp, which connects to the external rechargeable battery pack. The shape of the external rechargeable battery pack allows a user to align the battery with the shape of the nail lamp base for connecting the two portions together. When connected, where the lamp and battery portions meet, the exterior surfaces become flush with each other. There will be a seam that is between the nail lamp and the battery pack. At the seam, the surfaces of the lamp and battery are relatively flush with each other. The seam line remains visible and can be felt tactilely. The battery pack base plate can have a finger plate integrated with the plate. In an implementation, the finger plate is removable from the base plate to allow for replacement or cleaning between uses. More discussion on a finger plate is in U.S. provisional patent application 62/002,763, which is incorporated by reference. FIG. 20 shows a block diagram of a specific implementation of a kit 2000 for a nail lamp. The kit includes a UV light unit 2002, a battery pack 2004, a USB charger 2006, a USB charging cable 2008, and a user guide 2010 or instructions on use. These components can be arranged in a packaging of the kit which can include a box. In an implementation, the box can have compartments or trays for holding the components in place within the box. For example, one kit implementation is the system described in connection with FIG. 14 above. This kit has the battery pack connecting to the lamp with the USB connector input, and also the recharging circuitry is contained within the battery pack housing. Another kit implementation is the system described in connection with FIGS. 15-19 above. This kit has the battery pack directly connecting to the lamp, rather than through the USB connector input. The recharging circuitry is contained within the nail lamp housing. FIG. 21-23 show views of another implementation of a nail lamp 2100. FIG. 21 shows a perspective view, FIG. 22 shows a top view, and FIG. 23 shows a right side view. The nail lamp device has an exterior surface and at one side, an opening through which a user can place their hand into an interior space of the nail lamp. There are controls on the exterior that are used to turn on an interior lighting source of the device, which exposes the interior space to light from the interior lighting source. As an example, a user can insert their fingers into the interior space, turn on the cure interior lighting source, and cure their UV nail polish or UV nail gel coated nails with the interior light. In an implementation, the device includes sensors that detect when a hand is present inside the unit. This turns on both the interior curing lights as well as the exterior glowing lights for an allotted time (e.g., turning off after 15, 30, or 60 seconds). The light can also be manually turned on or off with, for example, button controls as an additional convenience. In an implementation, there is also an exterior lighting source of the device, which also turns on in response to the controls and is on when the interior lighting source is on. Light from the exterior lighting source is visible through a translucent shell (e.g., translucent plastic) of the exterior of the device. The translucent shell can be clear material or a light-diffusing material. When the interior lighting source is off, the light from the exterior lighting source will also be off. The exterior lighting source is used as an indicator that the device is on—that the interior lighting source is on. The entire exterior surface of the device can be lighted when on. This exterior lighting feature will make it easier for the user to know that the light is on and the curing cycle is continuing. The user will be able to see the exterior light is on from many positions and many angles, especially compared to attempting to peek into the opening (which will be partially blocked by a hand) and trying to see whether the interior lighting source is on. And the interior lighting source may not be visible light. In an implementation, on the exterior, there is a digital display. The display shows a length time in digits that the light will be turned on for. Further, the display can be a count down (or count up) timer that shows the time remaining for the light to be on. The digital display is optional and can be omitted in some implementations. More specifically, the nail lamp includes a housing 2102. The housing includes a base 2103 and an outer cover 2105. On a front side of the housing, there is an opening 2107 to a space (or cavity) within the housing. The space within the housing is defined by inner surfaces of the housing including a platform 2109, an inner side wall 2111, and an inner roof (not visible). The inner surfaces of the inside of the housing can be made of metal, plastic, or a combination of these. In an implementation, the opening is shaped and sized to allow a user's hand to pass through the opening into the space within the housing. The user's hand can be positioned within a cavity formed by the space, surrounded by the inner surfaces of the housing. In another implementation, the opening is adapted to allow a foot to pass through the opening. In another implementation, the nail lamp is adapted to be used for both a hand and foot. The outer cover of the housing includes a screen or display 2120 and controls, which in an implementation, are button features 2122a, 2122b, and 2122c. The screen may be an LED-backlit liquid crystal display (LCD) to display to a user a status or parameter of the nail lamp such as a time elapsed or a time remaining for a particular cure setting of the lamp. The display can also indicate other parameters of the lamp such as a power setting (e.g., “ON,” “OFF,” “LOW,” “HIGH,” or other messages). The screen can display images such as words, digits, 7-segment displays, meters, and others. The button features can indicate various cure settings of the nail lamp. Each button can be associated with a certain time of curing. For example, a first button can indicate a first timer setting for a first interval of time (e.g., 15 seconds). When a user selects the first timer setting by pushing the first button, a LED light source of the lamp will turn on for a time of 15 seconds of curing. A second button can indicate a second timer setting for a second interval of time (e.g., 30 seconds), and a third button can indicate a third timer setting for a third interval of time (e.g., 60 seconds). In other implementations, there can be fewer buttons (e.g., 1 or 2 buttons) or more than 3 buttons (e.g., 4, 5, or 6, or greater). FIG. 24 shows a view of an inside of a housing of a nail lamp, as viewed from a lower surface of the interior space looking toward the upper surface (e.g., inner roof). Side surfaces or side surfaces are angled with respect to the lower surface. The upper surface and side surfaces include a number of light source structures as shown. In an implementation, the light source structures are surface mounted light emitting diodes (LEDs). The LEDs can be referred to a surface mounted devices or SMDs. The LEDs are surface mounted to one or more printed circuit boards that housed within the device's enclosure, between surfaces of the interior space and exterior shell of the device. In other implementation, light sources can include other types of LEDs (other than SMDs), laser diodes, light bulbs, or other lighting. Some light source structures can be different from other light source structures. For example, first light structures 2421, 2423, 2425, 2427, 2429, 2431, 2433, 2435, 2437, 2439, 2441, 2443, 2445, and 2447 are different from the other light structures, which can be referred to as second light structures. In an implementation, the first light structures have higher energy output than the first light structures. For example, the first light structures can be 2-watt LEDs, while the second light structures are 1-watt LEDs. The light sources can include lights of the same or different output power and wavelength. In the specific arrangement of lights in FIG. 24, LED lights are positioned on the side walls and roof of the inside of the housing. There are seven side walls connected to the roof. The shaded LED lights (2421, 2423, 2425, 2427, 2429, 2431, 2433, 2435, 2437, 2439, 2441, 2443, 2445, and 2447) indicate 2-Watt output LEDs, while the remaining unshaded LED lights are 1-Watt output LEDs. Generally, on side walls of the housing, each 2-Watt LED is positioned between two 1-Watt LEDs. This distribution of LEDs can provide each nail of a user's hand (or foot) with an even exposure of light since a 2-Watt LED is positioned near each nail, as shown in FIG. 18. In other implementations, the LEDs can be arranged in another arrangement, such as an alternating pattern. On the inner roof of the housing, there is a combination of 2-Watt and 1-Watt LED lights. The 2-Watt LEDs can be arranged to correspond to a user's nails, so that a 2-Watt LED is near each nail. For example, when the user's left hand is inserted into a cavity of the housing, as shown in FIG. 18, each nail of the hand is irradiated by at least two nearby 2-Watt LEDs. Referring back to FIG. 4, with the user's hand placed in the cavity, each nail is irradiated by at least one nearby sidewall LED and one nearby inner roof LED. Table A below shows how each nail is irradiated for both right and left hands of the user. TABLE ARight HandLeft HandSidewallRoofSidewallRoofFingerLEDLEDFingerLEDLEDThumb nail421435Thumb nail433447Index nail425439Index nail429443Middle nail427441Middle nail427441Ring nail429443Ring nail425439Little nail431445Little nail423437 Each nail is also irradiated by at least two 1-Watt LEDs. For example, when the left hand is placed in the cavity, the thumbnail is irradiated by 2-Watt LEDs 2421 and 2437, and by the two 1-Watt LEDs surrounding LED 421. The index fingernail is irradiated by 2-Watt LEDs 2425 and 2439, and by two 1-Watt LEDs between LEDs 2425 and 2427, and between LEDs 2439 and 2441. FIG. 25 shows an inside view of a housing of a nail lamp in relief Light sources are positioned along sidewalls and inner roof of the housing. The side walls and roof include openings or apertures to expose a light source, which can be positioned in or behind the opening. Light from the light source radiates through the opening and into the space provided by the housing. By using surface mounted LEDs, the LEDs are recessed in openings of the enclosure. This is in comparison to other not-surface-mounted types of LEDs that have a bulb-portion that extend through the openings. Also in some implementations, the LEDs can be flush with the enclosure surface. FIG. 26 shows specific arrangement of LED lights on sidewalls and inner roof of a housing. The LEDs that are circled are 2-Watt LEDs using surface mount technology (SMT). These LEDs are referred to as surface mount devices (SMD) LEDs. The LEDs that are not circled, that are positioned between the 2-Watt LEDs, are 1-Watt SMD LEDs. In an implementation, a SMD LED can produce UV light in a range of about 340 nanometers to about 410 nanometers. In a specific implementation, the SMD LEDs can produce UV light at about 395 nanometers peak irradiance. In another specific implementation, the SMD LEDs can produce UV light at about 350 nanometers. In another specific implementation, the SMD LEDs can produce UV light at about 365 nanometers. FIG. 27 shows a specific arrangement of LED lights on sidewalls and inner roof of a housing with five inner sidewalls of the housing. The configuration of LED lights in FIG. 27 is slightly different from that shown in FIGS. 24, 25, and 26. There are two fewer LEDs than the other configurations. The circled LEDs indicate 2-Watt SMD LEDs, and the uncircled LEDs indicate 1-Watt SMD LEDs. For each sidewall, one 2-Watt LED is positioned between two 1-Watt LEDs. FIG. 28 shows a specific arrangement of SMD LED lights on sidewalls and inner roof of a housing with seven inner sidewalls of the housing. Compared to the arrangement in FIG. 7, this housing includes 2 additional sidewalls, each with a 2-Watt LED 2806 and 2808. So, the arrangement in FIG. 7 has five 2-Watt LEDs on sidewalls, while this arrangement includes seven 2-Watt LEDs positioned on sidewalls. The arrangement with two additional LEDs can increase the cost of the device, but provides the irradiation for curing, which can reduce curing time and improve a uniformity of the curing. FIG. 29 shows a top view of a finger plate 2901. The finger plate is placed onto the lower surface of the interior space of a nail lamp. The finger plate is a guide for the fingers, so the fingers will be properly positioned inside the nail lamp. The user places the fingers on the finger plate, and the nails are held in position for exposure to the curing light. The finger plate can be removable (e.g., sliding out from a bottom of the lamp), such as for cleaning or so other finger plates can be used for different sized fingers. The finger plate is designed for the right or left hand, but in other implementations, there may be a specific finger plate design for each hand. The finger plate includes five side by side depressions or grooves that are adapted to support a user's fingers when the user places a hand inside the housing on the plate. A first depression 2902 can be a sloped surface (or indentation, groove, or recess) for supporting the user's thumb or little finger. A second depression 2903 can be a groove (or indentation or recess) for supporting the user's index or ring finger. A third depression 2904 can be a groove (or indentation or recess) for supporting the user's middle finger. A fourth depression 2905 can be a groove (or indentation or recess) for supporting the user's index or ring finger. A fifth depression 2906 can be a sloped surface (or groove, indentation, or recess) for supporting the user's thumb or little finger. The finger plate can include thumb guides 2910 and 2911 that include circular grooves in the finger plate. The circular groove can provide a tactile guide for the user to place the thumb when the user inserts the hand into the housing. The thumb guide allows the user to keep the hand in the same position through the curing so that the nails cure evenly and without smudging. In an implementation, the finger plate is removable from the housing. Different finger plates can be used for users with different size hands. The finger plate can also be removed to facilitate cleaning of the plate and of the inside of the housing. In salons, the plate can be removed between uses to sterilize the plate for a new user. The finger plate can also be replaced with a foot plate for curing polish on a person's foot for a pedicure. FIG. 30 shows an outline of the finger plate overlaid on a bottom up view of an inside of a housing of a nail lamp. This figure shows the positioning of the light structures in relation to the finger plate grooves. Light sources are arranged along an inner roof of the housing. The roof includes openings or apertures to expose a light source (e.g., LED, or SMD LED, or others), which can be positioned in or behind the opening. Light from the light source radiates through the opening and into the space provided by the housing. FIG. 30 shows a specific arrangement of light sources relative to a finger plate of the housing. The finger plate includes finger grooves, with spacers (e.g., raised regions or ridges) between adjacent finger grooves. There is at least one light source positioned over each finger groove. Over a first finger groove 3002, there are two openings with a light source at each opening. There is a light source positioned over a second finger groove 3003, third finger groove 3004, and fourth finger groove 3005. A light source is positioned between and over the second and third finger grooves, and the third and fourth finger grooves. There are two light sources positioned over a fifth finger groove 3006. FIG. 31 shows a specific implementation of a finger plate 3101 with extended grooves for fingers of a user's hand. There can be spacers 3105 between adjacent grooves. The finger plate includes stops 3107 in some grooves to prevent the user's fingers from sliding in the grooves (e.g., away from or toward the light sources). The stops can provide a tactile gauge for the user to indicate where to place the fingers during curing. In a specific implementation, a height of the stops is about 3 millimeters from a surface of the groove. In other implementations, the height is less than 3 millimeters (e.g., 0.5, 1, 1.5, 2, or 2.5 millimeters or greater). In other implementations, the height is greater than 3 millimeters (e.g., about 3.1, 3.2, 3.3, 3.4, 3.5, 3.6, 3.7, 3.8, 3.9, or 4 millimeters or more). In an implementation, a finger plate can have shorter or longer grooves than that of FIG. 31. FIG. 32 shows an implementation of a finger plate with grooves that are shorter compared to the finger plate in FIG. 31. An edge 3202 of the finger plate provides a stop for a user's fingers. The edge can have raised regions or stops to provide the user with a tactile guide for placement of the fingers or fingertips. In a specific implementation, a height of the stops is about 1.5 millimeters from a surface of the groove. In other implementations, the height is less than 1.5 millimeters (e.g., 0.5, 1, 1.1, 1.2, 1.3, or 1.4 millimeters). In other implementations, the height is greater than 1.5 millimeters (e.g., about 1.6, 1.7, 1.8, 1.9, or 2 millimeters or more). In other implementations, the edge does not have a raised rim, and the user can place the fingertips at the edge itself. FIG. 33 shows the positioning of a user's hand (e.g., left hand) in the finger plate of FIG. 31, against the finger stops. FIG. 34 shows the positioning of a user's hand (e.g., left hand) in the finger plate of FIG. 32, against the finger stops. FIG. 35 shows a rear perspective view of a finger plate. A top view of the finger plate is in FIG. 29. As discussed, the plate can include five depressed regions (e.g., finger grooves) with adjacent regions separated by a raised region 3505 (or ridge). Three of the finger grooves, in the middle, are elevated compared to the other two finger grooves, on either side of the middle three. The depressed regions can be contoured or curved to provide comfort to a user's fingers when resting in the depressed regions. The depressed regions and raised regions can also prevent the fingers from moving while curing which can cause uneven curing or smudging. FIG. 36 shows a front perspective view of a finger plate. A first groove 3602 and a fifth groove 3603 are less raised from a base of the housing than second, third, and fourth grooves 3604, 3605, and 3606. The first and fifth grooves are slightly angled away from the second, third, and fourth grooves. A surface of the fingerplate between a front edge of the grooves and a base of the finger plate can be sloped. By elevating the second, third, and fourth finger grooves, the fingers will be positioned closer to the upper surface and the light structures. This will increase the radiation to the fingers which improve curing of the polish or gel. Curing time will be reduced and the uniformity of the curing will improve. Further, this structure reflects a natural positioning of a person's fingers at rest. So, when a user places fingers into the grooves of the finger plate, the fingers can rest in a natural position that ergonomic and comfortable than if the grooves were positioned at the same height from the base of the housing. FIG. 37 shows an irradiation pattern for light structures for the arrangement of FIG. 27. This specific arrangement of lights (e.g., LEDs) has sidewalls and inner roof of a housing with five inner sidewalls of the housing. A user's hand is positioned in the housing and each nail is irradiated by nearby light sources. A thumbnail is irradiated by three nearby light sources while a little finger nail 3705 is irradiated by two nearby light sources. In a specific implementation, for each sidewall of the housing, there is one 2-Watt LED that is surrounded by two 1-Watt LEDs. The thumbnail is irradiated by all three LEDs, while the little finger nail is irradiated by two 1-Watt LEDs. FIG. 38 shows an irradiation pattern for light structures for the arrangement of FIGS. 24, 25, 26, and 28. This specific arrangement of lights (e.g., LEDs) has sidewalls and inner roof of a housing with seven inner sidewalls of the housing. Compared to the arrangement in FIG. 37, there are two additional sidewalls 3803 and 3805, each sidewall with a light source 3806 and 3808. In this arrangement, the user's nails (right hand or left hand) can be evenly irradiated. The thumbnail and little finger nail of each hand can be each irradiated by at least three light sources. In a specific implementation, for each sidewall of the housing with three light sources, there is one 2-Watt LED that is surrounded by two 1-Watt LEDs. On each of sidewalls 3803 and 3805, there is one 2-Watt LED. The thumbnail and little finger nail is each irradiated by one 2-Watt LED and two 1-Watt LEDs. FIG. 39 shows a finger plate for an inside space having five inner sidewalls, such as used in connection with the light structure arrangement of FIG. 27. FIG. 40 shows a finger plate for an inside space having seven inner sidewalls, such as used in connection with the light structure arrangement of FIG. 28. The finger plates described in this application can be adapted or modified to be used with the configuration of FIG. 27 or 28, or both. For example, the finger plate in FIG. 40 can be used with the FIG. 27 configuration. And the finger plate in FIG. 39 can be used with the FIG. 28 configuration. Compared to the configuration in FIG. 39, two additional side walls 4006 and 4008 can be added at corners 3906 and 3908. The finger plate also includes indicator members 4010 (finger points) positioned in the grooves of the finger plate. In an implementation, the indicator members are raised dots or bumps analogous to Braille dots that provide the user a tactile guide that the fingertips are positioned properly. Note that for the first and fifth grooves, these include two indicator dots. This is because there grooves, depending on which hand, are for the thumb or pinkie, which are a different length. In other implementation, the indicator members can be other raised regions (e.g., bump, projection, or ridge, or others) or recessed regions that can provide the user tactile feedback. When the user inserts the hand into grooves of the finger plate, the user cannot see how far to extend the fingers into housing. With the indicator members, the user can feel where to position the hand during curing. FIG. 41 shows a front view of an inside of a housing of a nail lamp with an outer cover of the housing removed. The side walls and roof include openings 4105. Light source structures 4110 can be located in or behind the openings and are exposed through the openings. Light sources can be connected to circuit boards 4115. In a specific implementation, light sources are SMD LEDs that are mounted onto circuit boards. Circuit boards 4115 may be printed circuit boards upon which the surface mounted LEDs are soldered. There can also be heat sinks or heat fins to which the LEDs are attached to dissipate heat. There can be LEDs mounted on both sides of a printed circuit board. One side will include the LEDs facing the inside of the interior space, while the other side will include the LEDs for lighting the exterior of the device. There can be multiple printed circuit boards, with boards for the sidewalls and upper surface of the interior space. FIG. 42 shows a front view of an inside of a housing of a nail lamp with five inside side walls. Side walls are angled with respect to a vertical y-axis to allow the light sources to be angled toward a finger plate of the housing. In a specific implementation, an angle at which a side wall is angled with respect to the vertical axis is about 30 degrees. In other implementations, the angle is less than 30 degrees (e.g., about 20, 21, 22, 23, 24, 25, 26, 27, 28, or 29 degrees). In other implementations, the angle is greater than 30 degrees (e.g., about 31, 32, 33, 34, 35, 36, 37, 88, or 39 degrees, or more). FIG. 43 shows a front view of an inside of a housing of a nail lamp with seven inside side walls. Compared to the configuration in FIG. 42, the side walls can be less angled with respect to the vertical y-axis. In a specific implementation, an angle at which a side wall is angled with respect to the vertical axis is about 26 degrees. In other implementations, the angle is less than 26 degrees (e.g., about 18, 19, 20, 21, 22, 23, 24, or 25 degrees). In other implementations, the angle is greater than 26 degrees (e.g., about 27, 28, 29, 30, 31, 32, 33, 34, 35, 36, 37, 88, or 39 degrees, or more). FIG. 44 shows a top view of an exterior of a nail lamp. There are preset settings for a user to select for curing. In an implementation, the user can select a preset curing time (e.g., 15 seconds, 30 seconds, or 60 seconds). The UV nail lamp in FIG. 44 is set to a setting of 60 seconds curing time. When the user presses the button for the selected setting, the button can light up and remain lit during the curing. A display can indicate to the user how much time has elapsed or is remaining on the curing time. The display shows 20 seconds (or 2 seconds) has elapsed or is remaining of the selected 60 seconds. Once the time expires, the UV lights, along with the lights of the housing, will turn off. In an implementation, when the user selects the desired cure time by pressing the button, the display will display the selected time. In an implementation, an exterior lighting source of the device does not turn on until a person's hand is inserted inside of the nail lamp. When the hand is inside, a sensor of the device detects when a hand is present inside the unit. This turns on both the interior curing lights as well as the exterior glowing lights for duration of the selected curing. When curing begins, exterior light source of the device will turn on, causing the exterior surface of the lamp to glow a soft and steady light for the duration of the curing time. The exterior lights can be positioned within the device, between interior curing lights and an outer translucent cover of the device. The translucent cover can be a translucent plastic material. The translucent plastic material can be a diffusing material or a diffuser, or the translucent plastic material can be combined with another diffusing material or diffuser, such as a composite material including both a translucent plastic layer and a light diffusing layer. In an implementation, the translucent plastic material of the lamp shell includes a light diffusing property. When light irradiated from the exterior light source hits an inside surface of and is transmitted through the translucent plastic material, the plastic material diffuses or spreads out (i.e., scatters) the light to give a softer light relative to the more concentrated light initially radiated from the exterior lighting source (e.g., diode on the circuit board). The scattered light can be across the entire exterior shell and cause the device to have a soft and steady glow of light. For example, in FIG. 44, about six exterior lights sources are used to illuminate and cause the lamp's exterior surface to glow. The light diffuser material spreads and homogenizes the non-uniform or uneven illumination of six light sources into a more uniform illumination. In an implementation, light diffusing property is present across an entire exterior surface area of the shell. When light from an exterior lighting source (located inside the nail lamp housing) enters an inside surface of the lamp shell, the light diffusing material scatters the light across the entire exterior surface area of the shell. This causes a more even glow across the entire lamp shell. In an implementation, the lamp shell has a light diffusing property when the lamp shell is made of a translucent material and a light diffuser film is coupled to an interior surface, or exterior surface, or both interior and exterior surfaces of the translucent lamp shell material. Examples of light diffusing films includes mylar or acetate, or similar films. Other examples of light diffusing film include films that have varying degrees of opacity. In another implementation, the lamp shell has a light diffusing property when the lamp shell includes a roughened surface, which scatters light. In a specific implementation, the lamp shell includes randomly sized and randomly placed particles on a surface of the lamp shell. In another specific implementation, particles can be of sizes large enough to be visible to the eye. In another specific implementation, the lamp shell includes a matting agent. The matting agent can blur spots of relatively more intense light produced by individual light sources. Examples of a matting agent can include silica powder, calcium carbonate powder, alumina powder, or the like. In a further implementation, the matting agents can have a particle size of approximately 1 to 5 microns. In an implementation, the light diffusing material is positioned over all of the exterior lighting sources so that all of the light from the exterior lighting sources will enter the light diffusing material and exit as an even glow that is spread across the entire surface of the shell. In a specific implementation, the light diffusing material is applied over an entire inner surface of the shell. In another implementation, the light diffusing material is applied over an outer surface of the shell. In another implementation, the light diffusing material is positioned over a portion of the exterior lighting sources. A portion of the light will enter and exit the light diffusing material and a portion of the light will not enter the light diffusing layer. This can result in various glow patterns across the shell the nail lamp. Each glow pattern can have a functional purpose, such as using a certain glow pattern to show when customers are close to finishing curing their gel nail polishes. In an implementation, a greater portion of the lamp shell's exterior surface area includes light diffusing property (or light diffusing material) than a portion that does not have light diffusing property. In another implementation, the lamp shell's exterior surface includes a portion with light diffusing property and an opaque portion, which does not let light travel through. In a specific implementation, the portion of the lamp shell's exterior surface that includes light diffusing property ranges from 10%-100%. The remaining portion of the lamp shell's exterior surface is opaque. In another implementation, the lamp shell's exterior surface includes a portion with light diffusing property, a transparent portion, and an opaque portion. In an implementation, the nail lamp housing includes a first layer with light diffusing properties that is coupled to a second layer of material, which blocks out light. In a specific implementation, the light blocking material can block out specific wavelengths of light, such as UV light. Some of the interior light sources can emit UV light. Though the interior light sources are directed into the cavity (or interior space), some light rays may reflect off the inner walls of the cavity and be emitted through the shell of the nail lamp. To prevent the UV light from emitting through the shell, a layer of UV light blocking material can be added to the housing. Examples of materials that block out UV light are polycarbonate, acrylic, acrylic glass, and the like. In an implementation, the exterior light sources are positioned in regions of rather than the entire device. For example, the exterior lights can be positioned along an outer perimeter of the device. When the light is transmitted through and scattered by the translucent outer cover, the regions closest to the light sources will glow brighter than the regions farther away from the light sources (e.g., a top region of the outer cover). Typically, the LEDs for the exterior lighting are not the same wavelength as the interior lighting. In an implementation, the exterior lights are non-UV lights. In an implementation, these lights can produce visible colored light, all the same color, such as in blue. Other colors can include pink, orange, yellow, red, green, or purple or others. In other implementations, there can be different colors of exterior light (such as blue and yellow, or red and green). In other implementations, the lights are LEDs such as RGB LEDs that can produce changing colors of light during curing. FIG. 45 shows a perspective view of an exterior of a nail lamp. The display shows 44 seconds has elapsed or is remaining of the selected 60 seconds. Once the time expires, the UV lights, along with the lights of the housing, will turn off. FIG. 46 shows a top perspective view of an exterior of a nail lamp that is turned on (i.e., curing mode). A timer displays 20 seconds (or 2 seconds) has elapsed or is remaining of the selected 60 seconds. UV lights on an inside of the housing are turned on, and glow from an opening of the housing of the lamp. A specific process flow for operating a UV nail lamp is presented in Table B below. It should be understood that the invention is not limited to the specific flows and steps presented. A flow of the invention may have additional steps (not necessarily described in this application), different steps which replace some of the steps presented, fewer steps or a subset of the steps presented, or steps in a different order than presented, or any combination of these. Further, the steps in other implementations of the invention may not be exactly the same as the steps presented and may be modified or altered as appropriate for a particular application. TABLE BStepFlow1Power on UV lamp.2Select curing mode. This can include a user selecting a curingtime, or a level of curing, or other parameters from a presetoptions (e.g., menu or buttons). The user can also manually inputa desired curing time or level of curing (e.g., buttons, dial,knob, or menu). In an implementation, the user presses oneof a plurality of buttons to select a predetermined curingtime (e.g., 15 seconds, 30, seconds, and 60 seconds).A display can display the selected curing time or setting.Lights between an inside of the housing and an outercover of the housing will light up, causing the housing to lightup or glow during curing.3A user inserts a hand (or foot) into the housing. The user'shand can rest on a finger plate. The finger plate can havefinger indicator members that allow the user to feel whereto rest the fingertips.4Timer starts when the user's hand is inside the housing. As thetimer starts, UV light sources within the housing turn on toirradiate the user's nails.5Timer stops after the selected time expires. When the timerstops, the UV light sources turn off. Lights between the insideof the housing and the outer cover of the housing will turnoff, causing the housing to dim.6User removes hand from the housing.7Power off UV lamp. FIG. 47 shows a block diagram of a specific implementation a nail lamp that is adapted to be used with a power source that is external to the nail lamp. The nail lamp includes a shell 4702 (also referred to as an exterior surface) and an enclosure 4704 (also referred to as a cavity or interior space), which is defined by an upper surface 4706 (also referred to as inner wall of a nail lamp's housing) of the enclosure. A user can place a hand inside the enclosure. A removable finger plate 4708 can optionally attach to the nail lamp and further define the enclosure. A power circuit 4710, inside the lamp, is coupled to an external battery 4712 or an adapter 4714, both of which are outside of the nail lamp. The external battery can be connected to a charger 4716. The adapter can be connected to an external power supply (e.g., a wall outlet). The external battery or external power supply provides power to a power circuit. The power circuit provides power to sensors 4718, one or more interior LEDs 4720, a control circuit 4722 that includes a control unit 4722 and a timer display 4726, and one or more LED units 4728 that include exterior LEDs 4730 and interior LEDs 4720. The interior LED can also be referred to as an interior lighting source, discussed above, and used to cure the gel polish. The exterior LED can also be referred to as an exterior lighting source, discussed above, and produces light to indicate that the interior LED is activated. A button 4732, located outside of the shell, is connected to the control circuit. When pressed, the button activates the control circuit that controls the timer display and activates one or more SMD interior LEDs 4720 or LED units 4728. Heat sinks can be coupled to the interior LEDs within the shell. The heat sink can absorb heat given off by an activated LED so that a user's hand will not feel hot and uncomfortable inside the nail lamp. The power circuit can optionally include an internal battery 4734. The internal battery can be charged by connecting to an external battery or an adapter that is connected to an external power source such as a wall outlet. After the internal battery has been charged by the external battery or external power supply, the nail lamp can operate without being connected to an external battery or adapter. The power circuit can also include a switch between the internal battery and external power connections (e.g., such as connection to an external battery or wall outlet) to allow the nail lamp to switch between internal and external power sources. FIGS. 48-50 show an implementation of a nail lamp 4802 that includes a battery input port 4804 (also referred to as a power input) so that the nail lamp can be used with a rechargeable battery pack that is external to the housing of the nail lamp. The rechargeable external battery 4806 can provide power to the nail lamp. The external battery can be removably coupled to a cable 4808, which is removably coupled to the battery input port. FIG. 48 shows a block diagram of nail lamp 4802. FIG. 49 shows a side view of the nail lamp including the external battery attached to the nail lamp via the cable. FIG. 50A shows a first short side of the external battery. FIG. 50B shows a second short side of the external battery. FIG. 50C shows a first long side of the external battery. FIG. 50D shows a top face of the external battery. The external battery supplies power to the nail lamp. With an external battery coupled to the nail lamp and providing power, the nail lamp does not have to be coupled to a wall outlet or laptop for power supply, the nail lamp can be moved around a room to any location. To charge the external battery, the external battery can be connected to an adapter, which can be connected to a wall outlet. The external battery can also be charged by being connected to a charging dock. After the external battery is charged, it can be disconnected from the adapter or dock and coupled to the nail lamp. FIG. 51 shows a block diagram of a charging dock 5102 and an external battery 5104. The charging dock includes a battery dock 5106 for the external battery, and optionally a latch 5108 to prevent the battery from falling out of position in the battery dock. Once the external battery is inserted into the battery dock, the charging dock starts charging it. The charging dock stops charging the external battery after the battery is removed. The charging dock can be connected to a power supply via a cable 5110 that can be connected to an adapter 5112, which can be connected to the power supply (e.g., a wall outlet). FIGS. 52-54 show an implementation of a nail lamp 5202 including a battery dock attachment 5204 that can be removably coupled to an exterior of the nail lamp. FIG. 52 shows a block diagram of the nail lamp and the battery dock attachment. FIG. 53 shows a side view of the nail lamp and the battery dock attachment attached to the nail lamp. FIG. 54 shows a side view of the nail lamp with the battery dock attachment detached from the nail lamp. The battery dock includes a slot for a battery 5208 and a latch 5210 to hold the battery firmly to the battery dock. The latch can be, for example, a spring loaded release latch. The battery can be inserted into the slot. The battery dock attachment provides for easy removal of the battery when the battery needs to be recharged. FIGS. 55-57 show an implementation of a nail lamp 5502 that includes an internal battery dock 5504 where a rechargeable battery pack 5506 can integrate with the housing of the nail lamp. The internal battery dock is removably coupled to a battery 5506 to be removably coupled within the housing of the nail lamp. FIG. 55 shows a block diagram of the nail lamp including the internal battery dock. FIG. 56 shows a specific implementation of nail lamp 5502 in which the internal battery dock is located at a bottom 5606 of the nail lamp. The battery can be inserted into the bottom of the nail lamp. In other implementations, the battery dock can be located elsewhere, such as the top or side of the nail lamp, for easy access to the battery dock. The internal battery dock optionally includes a latch 5508 to hold the battery firmly to the battery dock. The latch can be, for example, a spring loaded release latch. The battery can be inserted into the slot. FIG. 57 shows a perspective view of the battery. The battery can include leads (e.g., copper strips) or pins that interface with the battery dock. FIG. 58 shows a specific implementation of an interior lighting source unit 5801. The interior lighting source unit includes at least one UV wavelength (which is approximately 100-400 nanometers) light source and at least one LED. The LED can produce light of a wavelength that is same or different from that produced by a UV wavelength light source. In a specific implementation (shown in FIG. 59), four UV light sources and one LED can be arranged such that the one LED lighting source 5803 is in the middle and the UV light sources 5805 surround the LED lighting source on four sides, like a rectangle, or square, or diamond shape. FIG. 59 shows another arrangement 5901 where three UV lighting sources surround one LED lighting source in a triangle shape. In a specific implementation, the LED produces light of 405 nanometers and can be 1-3 Watt LEDs. In another specific implementation, the UV lighting source produces light of 365 nanometers. FIG. 60 shows a strip 6001 of interior lighting source units 6002 and a magnification (indicated by broken line 6003) of one of the interior lighting source unit. An LED 6004 is adjacent to another LED 6006. The LEDs produce light of different wavelengths from each other. In a specific implementation, LED 6004 produces light of 405 nanometers, which can be used to cure LED gel. And LED 6006 produces light of 365 nanometers, which can be used to cure UV curable gel or extension gel. This arrangement of UV and LED light sources allow for universal usage of the nail lamp because the nail lamp can be used to cure both LED and UV-curable gel polish. In a further implementation, the nail lamp can be an inductive nail lamp, which the power required to generate light is transferred from outside the nail lamp to the gas inside via an electric or magnetic field. A benefit to an inductive nail lamp is extended lamp life. This description of the invention has been presented for the purposes of illustration and description. It is not intended to be exhaustive or to limit the invention to the precise form described, and many modifications and variations are possible in light of the teaching above. The embodiments were chosen and described in order to best explain the principles of the invention and its practical applications. This description will enable others skilled in the art to best utilize and practice the invention in various embodiments and with various modifications as are suited to a particular use. The scope of the invention is defined by the following claims. |
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048667447 | abstract | A scattering beam eliminating member is located between an X-ray tube and an X-ray gas detector such that it is intimately bonded to the window of the X-ray detector. The scattering beam eliminating member is of such a type that plates and X-ray transmission areas are alternately arranged in a slice direction to permit any scattering beam to be absorbed in the slice direction and never to be incident to the X-ray detector. |
abstract | An X-ray fluorescence analyzer includes: a measurement device having: an X-ray source that emits an X-ray; an irradiation area restricting member that restricts an area of a measurement sample to be irradiated with the X-ray as a primary X-ray; and a detector that detects a secondary X-ray generated from the measurement sample. The analyzer further includes: a sample stage that holds and moves the measurement sample between a measurement position at which the measurement sample is irradiated with the primary X-ray to detect the secondary X-ray by the detector and a first retracted position at which the measurement sample is retracted from the measurement position; and a calibration sample moving mechanism that holds a calibration sample for calibrating the measurement device and moves the calibration sample between the measurement position and a second retracted position at which the calibration sample is retracted from the measurement position. |
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052232115 | abstract | A zirconium based alloy plate of low irradiation growth, containing not more than 5 wt % Sn and/or not more 5 wt % Nb, and the balance Zr of not less than 90 wt %. The alloy plate has a texture that <0001> orientation (Fr value) with respect to direction perpendicular to the surface of the plate ranges from 0.20 to 0.50. This alloy plate is effective in being used to form a fuel channel box. Also a fuel assembly using this channel box is provided in which the crystal orientation of the zirconium alloy is made random by a heat treatment. Specifically, by the heat treatment, the Fr, Ft, and Fl values thereof are set to 0.25 to 0.50, 0.25 to 0.36, and 0.25 to 0.36, respectively. |
summary | ||
summary | ||
040382028 | claims | 1. In a process for the preparation of microspheres of an actinide metal oxide or actinide metal salt including dispersing an aqueous metal oxide or salt phase, or an aqueous metal oxide phase additionally containing carbon, onto an immiscible organic liquid under conditions such that the dispersed droplets are solidified into solid microspheres and separating the thus produced solid microspheres with some organic liquid adhering thereto, the improvement of removing said organic liquid from said microspheres comprising, in combination 2. washing the thus separated microspheres with an aqueous solution containing a solubilizing amount of a hydrophilic surface active agent having a HLB value of at least 10.0 up to about 14. 1. providing said organic liquid with a lipophilic surfactant therein having an HLB value from about 7 to less than about 10.0 2. The process as claimed in claim 1 wherein said organic liquid is composed of at least one alkyl substituted benzene having 8-20 carbon atoms in its alkyl chain. 3. The process as claimed in claim 1 wherein said aqueous washing solution contains a blend of alkyl aryl sulphonates and polyoxyethylene sorbitan esters of fatty and resin acids as said hydrophilic surface active agent. 4. The process as claimed in claim 1 wherein said organic liquid contains a particle dispersing amount of a blend of polyoxyethylene sorbitol fatty acids and alkyl aryl sulphonates as said lipophilic surface active agent. 5. The process as claimed in claim 1 wherein said aqueous metal oxide or salt phase is selected from the group consisting of actinide metal hydroxide sols, actinide metal salt solutions, anion deficient metal salt solutions and mixtures thereof. 6. The process as claimed in claim 5 wherein said aqueous metal oxide or salt phase additionally contains colloidal carbon distributed throughout said phase. 7. The process as claimed in claim 5 wherein said aqueous metal oxide or salt phase additionally contains at least one ammonia liberating agent. 8. The process as claimed in claim 1 wherein said organic liquid contains ammonia or an ammonia liberating agent. 9. The process as claimed in claim 1 wherein said aqueous metal oxide or salt phase contains at least one actinide metal and at least one element selected from the group consisting of Zr.sup.+.sup.4, Hf.sup.+.sup.4, Y.sup.+.sup.3, Sc.sup.+.sup.3 and rare earth elements admixed with said actinide metal oxide. 10. The process as claimed in claim 1 wherein said aqueous metal oxide or salt phase is selected from the group consisting of uranyl nitrate and anion-deficient uranyl nitrate. |
description | As shown in FIG. 1, a conventional fuel assembly 10 of a nuclear reactor, such as a Boiling Water Reactor (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 (also known as spacer grids) 15 axially spaced one from the other and maintaining the rods 18, 19 in the given matrix thereof. The fuel rods 18 and 19 are generally continuous from their base to terminal, which, in the case of the full length fuel rod 18, is from the lower tie plate 16 to the upper tie plate 14. Outer channel 12 encloses the fuel rods 18/19 within the assembly 10 and maintains water or other coolant flow within assembly 10 about fuel rods 18/19 and in contact with the fuel rods 18/19 to facilitate heat transfer from the fuel to the coolant. Outer channel 12 is traditionally uniform in mechanical design and material for each other assembly 10 provided to a particular core, to aid in assembly design standardization and manufacturing simplicity. Outer channel 12 may be fabricated conventionally of a material compatible with the operating nuclear reactor environment, such as a Zircaloy-2. As shown in FIG. 2, a conventional reactor core, such as a BWR core, may include a plurality of cells 40 in the reactor core. Each cell may include four fuel assemblies 10 having adjacent fuel channels 12. Other fuel assemblies 10 may be placed in the reactor core outside of cells 40 and not adjacent to control blades. The fuel assemblies 10 in FIG. 2 are shown in section to illustrate control blades 45, which are conventionally cruciform-shaped and movably-positioned between the adjacent surfaces of the fuel channels 12 in a cell 40 for purposes of controlling the reaction rate of the reactor core. Conventionally, there is one control blade 45 per cell 40. As a result, each fuel channel 12 has two sides adjacent to the control blade 45 and two sides with no adjacent control blade. The control blade 45 is formed of materials that are capable of absorbing neutrons without undergoing fission itself, for example, boron, hafnium, silver, indium, cadmium, or other elements having a sufficiently high capture cross section for neutrons. Thus, when the control blade 45 is moved between the adjacent surfaces of the fuel channels 12, the control blade 45 absorbs neutrons which would otherwise contribute to the fission reaction in the core. On the other hand, when the control blade 45 is moved out of the way, more neutrons will be allowed to contribute to the fission reaction in the core. Conventionally, only a fraction of all control blades 45 within a core will be exercised to control the fission reaction within the core during an operating cycle. As such, only a corresponding fraction of fuel assemblies will be directly adjacent to an extended control blade, or “subject to control,” during an operating cycle. After a period of time, a fuel channel 12 may become distorted as a result of differential irradiation growth, differential hydrogen absorption, and/or irradiation creep. Differential irradiation growth is caused by fluence gradients and results in fluence-gradient bow. Differential hydrogen absorption is a function of differential corrosion resulting from shadow corrosion on the channel sides adjacent to the control blades 45 and the percent of hydrogen liberated from the corrosion process that is absorbed into the fuel channel 12; this results in shadow corrosion-induced bow. Irradiation creep is caused by a pressure drop across the channel faces, which results in permanent distortion called creep bulge. As a result, the distortion (bow and bulge) of the fuel channel 12 may interfere with the movement of the control blade 45. Channel/control blade interference may cause uncertainty in control blade location, increased loads on reactor structural components, and decreased scram velocities. Conventionally, if channel/control blade interference has become severe, the control blade is declared inoperable and remains fully inserted. Example embodiments are directed to fuel assemblies useable in nuclear reactors and methods of optimizing and fabricating the same. Example embodiment fuel assemblies include an outer channel having a physical configuration determined based on a position of the fuel assembly within a core of the nuclear reactor, such as the position of the fuel assembly with respect to a control blade in the nuclear reactor that will be used to control core reactivity. When example embodiment fuel assemblies are to be directly adjacent to an inserted control blade, the outer channel may be thickened, reinforced, and/or fabricated of a material more resistant to deformation than Zircaloy-2, such as Zircaloy-4, NSF, and VB, so as to reduce or prevent distortion of the channel against the control blade and interfering with operation of the same. When example embodiment fuel assemblies are not in a controlled location, the outer channel may be thinned so as to increase water volume and reactivity in the assembly. As such, a reactor core including example embodiment fuel assemblies will include fuel assemblies having unique outer channels, in thickness, material, etc., unlike conventional power reactor cores. Example methods of configuring fuel assemblies include determining operational characteristics of the fuel assembly, such as the likelihood that the fuel assembly is controlled via control blade insertion in the nuclear reactor in a current or future fuel cycle, and physically selecting or modifying the outer channel of the fuel assembly based thereon. For example, if the fuel assembly is in a controlled location during the fuel cycle, the outer channel may be fabricated of a material more resistant to deformation than Zircaloy-2, such as Zircaloy-4, NSF, or VB, and/or thickened. Or, for example, if the fuel assembly is not in a controlled location, the outer channel may be approximately 20 mils (thousandths of an inch) or more thinner than outer channels of conventional fuel assemblies. Example methods are useable with or may further include configuring outer channel characteristics in order to meet desired neutronic properties of the fuel assembly. Hereinafter, example embodiments will be described in detail with reference to the attached drawings. However, specific structural and functional details disclosed herein are merely representative for purposes of describing example embodiments. The example embodiments may 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.). 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. As used herein, “channel,” “outer channel,” and the like are defined in accordance with the conventional fuel assembly structures shown and described in FIG. 1 as element 12, subject to the modifications discussed hereafter. As used herein, “distortion” or “channel distortion” includes both channel bow and channel bulge in nuclear fuel assemblies that may cause interference with control blade operation. It should also be noted that in some alternative implementations, the functions/acts noted may occur out of the order noted in the figures or described in the specification. For example, two figures or steps shown in succession may in fact be executed in parallel and concurrently or may sometimes be executed in the reverse order or repetitively, depending upon the functionality/acts involved. The inventors of the present application have identified several potential fuel channel characteristics and/or modifications to reduce or prevent fuel channel distortion. The inventors of the present application have further identified the effect these characteristics, in combination with other fuel assembly parameters, have on whole core performance. Example embodiments and methods discussed below uniquely address these previously-unidentified effects to achieve several advantages, including improved core performance, increased energy generation, reduced control blade error, materials conservation, and/or other advantages discussed below or not, in commercial nuclear power plants, while departing from total fuel channel uniformity purposefully used in conventional commercial nuclear power plants. Example embodiment fuel assemblies include fuel channels with optimized physical properties. Example embodiment fuel assemblies may include one or more channel characteristics to decrease fuel channel distortion. For example, the fuel channel may be thickened in its shortest dimension or reinforced with additional material. The thicker or reinforced fuel channel has greater resistance to distortion from differential irradiation growth, differential hydrogen absorption, and/or irradiation creep experienced in operating nuclear reactor environments. The percent reduction in deformation is approximately proportional the percentage increase in channel thickness. Or, for example, materials may be used in the channel that are resistant to distortion. For example, Zircaloy-4, a known zirconium alloy excluding nickel, may replace Zircaloy-2, which contains nickel. The reduced nickel content in Zircaloy-4 reduces differential hydrogen absorption and resultant channel bow. Other materials more resistant to deformation than Zircaloy-2 may additionally be used in whole or in part in addition to Zircaloy-4. For example, additional materials more resistant to deformation than Zircaloy-2 are described in co-pending application Ser. No. 12/153,415 “Multi-layer Fuel Channel and Method of Fabricating the Same,” incorporated herein by reference in its entirety. That document discloses alloys hereinafter called “NSF” having about 0.6-1.4% niobium (Nb), about 0.2-0.5% iron (Fe), and about 0.5-1.0% tin (Sn), with the balance being essentially zirconium (Zr) and alloys hereinafter called “VB” having about 0.4-0.6% tin (Sn), about 0.4-0.6% Fe, and about 0.8-1.2% chromium (Cr), with the balance being essentially zirconium (Zr). Other configurations for decreasing fuel channel distortion are useable with example embodiment fuel assemblies. Example embodiment fuel assemblies may use multiple mechanisms in combination to further reduce fuel channel distortion. Configurations and fuel channel characteristics in example embodiment fuel channels may be selected in accordance with example methods, discussed in the following section. Example embodiment fuel assemblies may further include channel characteristics that improve fuel neutronic characteristics, decrease material usage and costs, and/or improve other fuel assembly parameters. Such characteristics may include, for example, a thinner channel that permits greater water volume and neutron moderation within example embodiment fuel assemblies. The thinner channel may consume less material in fabrication and improve fuel assembly reactivity, heat transfer characteristics, etc. Example embodiment fuel assemblies having thicker, reinforced, and/or thinner channels, different alloys, or other channel modification may be used instead of conventional fuel assemblies having standardized channels throughout an entire core. Example embodiment fuel assemblies may thus significantly improve performance of a core including example embodiment fuel assemblies and/or reduce fuel resource consumption. For example, thinning the channels of 75% of the fresh conventional fuel assemblies for a particular fuel cycle by approximately 20 mils (20 thousandths of an inch) in the thinnest dimension may result in a reduction in volume of approximately 16,500 in3 zirconium alloy used. In the same example, assuming 8 channels would not need to be replaced because they include channel mechanisms to decrease fuel channel deformation, an additional ˜2,000 in3 zirconium alloy volume may be saved. In the same example, assuming 8 channels are not needed to be fabricated because 8 fewer fuel assemblies are required in a fuel cycle with fuel savings from channel characteristics that improve fuel neutronic characteristics of example embodiment fuel assemblies, an additional ˜2,000 in3 zirconium alloy volume may be conserved. Thus, example embodiment fuel assemblies, having different channel characteristics selected and implemented in accordance with example methods discussed below, may result in significant materials savings and improved core performance. As discussed above, increasing channel thickness decreases water volume and overall reactivity of an assembly having a thicker channel. Lower reactivity results in less optimal fuel usage and less power production in a nuclear core of a nuclear power reactor. Increasing channel thickness further increases costs of fuel assemblies having thicker channels. Increasing channel thickness also reduces the risk and/or magnitude of channel distortion and interference with control blade function. Decreasing channel thickness has a generally opposite effect of increasing water volume and overall reactivity of an assembly having a thinner channel, while also increasing distortion likelihood. Zircaloy-4 has similar fluence bow and creep bulge characteristics compared to Zircaloy-2. Zircaloy-4, however, resists channel bow caused by differential hydrogen absorption. NSF and VB are additionally resistant to other forms of bow and bulge causing channel deformation. Example methods uniquely leverage the above advantages and disadvantages of fuel channel modification to reduce or prevent channel distortion while minimizing negative effects on fuel economy, control blade function, and other core performance metrics. As shown in FIG. 3, example methods include an operation S100 of determining fuel assembly characteristics, including whether a fuel assembly is placed or will be located in a cell such that the fuel assembly will be directly adjacent to a control blade that will be operated in a current and/or future fuel cycle to control the fission reaction in the core. A fuel assembly positioned directly adjacent to a control blade that is likely to be exercised to control the fission reaction is herein defined as a “controlled fuel assembly” or in a “controlled location,” because it is most subject to control blade negative reactivity and most likely to affect control blade performance. The determination of whether a fuel assembly is subject to control may be based on one or more fuel assembly operational characteristics that determines placement/position of the fuel assembly within the reactor core over one or more fuel cycles, in addition to overall plant characteristics such as core size, thermal power rating, etc. For example, an operational characteristic may be reactivity of the fuel assembly. Reactivity determines the degree to which the fuel can contribute to the fission chain reaction during power operations. Reactivity is directly controllable with control blade insertion, due to the blades' neutron-absorbing properties. As such, fuel with higher reactivity may be placed in controlled locations to enhance core-wide overall control of the neutron chain reaction. Similarly, fuel with lower reactivity may be less likely to be subject to control. Although location with regard to utilized cruciform control blades is described in connection with example embodiments and methods, it is understood that other sources of negative reactivity may additionally be accounted for in example methods and embodiments. For example, proximity to burnable poisons or proximity to a control rod present in some plant designs may be accounted for by determining operational characteristics of the fuel assembly that determine the likelihood that the fuel assembly will be placed in that proximity. Controlled locations may also be determined in S100 by known core modeling and mapping methods and software. For example, a program may receive input of several fuel assembly operational characteristics for several fuel assemblies and determine an optimum core configuration with corresponding fuel assembly positions. Because example methods and embodiments may themselves affect fuel assembly operational characteristics as discussed below, such known core modeling and mapping methods may be alternatively and repetitively executed before and following fuel assembly modification in example methods to ensure optimized core performance. Following the determination in S100, one or more fuel assembly channels are configured based on the position determination. The configuring generally increases assembly reactivity, decreases distortion potential, and/or reduces material consumption in the configured assembly/assemblies. If it is determined from S100 that the assembly will be placed in a cell adjacent to an employed control blade, i.e., subject to control, then a first configuration S210 is pursued. S210 configures the assembly channel to reduce or eliminate channel distortion during power operations. For example, in S210, channel thickness may be increased by several hundredths of an inch or more to ensure decreased channel distortion. The degree of thickening may further be based on decreased reactivity or other operational characteristics desired of the assembly during operation in the nuclear reactor core. Or, channel thickness may be increased or the channel may be reinforced on only a side or wall directly adjacent to the control blade that will be operated, while remaining fuel channel sides may be unmodified or modified in accordance with S220. Additionally, or in the alternative, in S210, the channel may be fabricated out of a material more resistant to distortion than Zircaloy-2, including shadow-corrosion-bow-resistant Zircaloy-4, or fluence-gradient-bow and/or-creep-bulge-resistant NSF or VB. In this way, only assemblies determined to be at a position benefitting from a thicker or reinforced channel or a channel including Zircaloy-4, NSF, and/or VB, such as a controlled assembly likely to be placed in a cell adjacent to an employed control blade, are configured with channel features that decrease or eliminate distortion while leveraging other characteristics such as reactivity or fabrication expense. Further, because assemblies in a controlled core position typically possess higher excess reactivity, a thicker or reinforced channel that may decrease reactivity is not a significant disadvantage for the overall core reactivity; indeed, such reactivity-decreasing configuration may aid in balancing core power production and/or simplifying control blade operations. If it is determined from S100 that the assembly will be placed in an uncontrolled core position, such as an edge position in the core or adjacent to a control blade that will not be utilized, then a second configuration S220 is pursued. S220 configures the assembly channel to increase fuel assembly neutronic characteristics for the assembly in the operating core and decrease manufacturing burden in fabricating the assembly, without regard to distortion risk. For example, in S220, channel thickness may be decreased by several hundredths of an inch or more to increase water or moderator volume in the assembly, thereby increasing reactivity and fuel usage in the assembly. Reducing channel thickness in S220 further decreases an amount of expensive zirconium alloy or other channel material required to fabricate the assembly. In S220, assembly channel thickness may be reduced by a margin that takes into account the increased reactivity; the channel may be thinned such that the assembly has a determined or desired reactivity or other operational property when in use in the nuclear reactor core. In this way, a core may contain fuel assemblies with several different, unique channel thicknesses and other characteristics as determined in S210 and S220. Assemblies may be configured in S210 and S220 in several different manners and timeframes. For example, the configuring in S210 and S220 may be selecting a pre-existing assembly or ordering an assembly having the configuration determined in S210 and S220, by a power plant operator, for insertion or re-insertion during an upcoming fuel cycle in the nuclear reactor core. Alternatively, the configuring in S210 and S220 may be a physical fabricating or modifying of the fuel assembly to match the configuration determined in S210 and S220 by a fuel assembly manufacturer or refitter, for example. Example methods including S100 and S210/S220 may address fuel assembly location and configuration for use in an immediately approaching fuel cycle, a future fuel cycle, and/or multiple fuel cycles. For example, S100 may determine that a fuel assembly will be in a controlled position adjacent to an employed control blade in a first fuel cycle, and the same or later analysis may determine that the fuel assembly will be relocated to a position away from a control blade in a second layer fuel cycle. The assembly may be configured under S210 for the first cycle, and then reconfigured under S220 for the second cycle. Such reconfiguring may include re-channeling the fuel assembly by removing and replacing the channel used in the first fuel cycle with a channel having the configuration determined in S220 for use in the second fuel cycle. Similarly, a reverse determination may result in the reverse configuration. Or, for example, S100 may determine, based on multi-cycle operating parameters, that a particular fuel assembly will not be placed in a controlled location in its lifetime. Configuration of the assembly may then proceed under S220, without further modification of the assembly during its lifetime in the reactor. FIG. 4 is an illustration of an example reactor core 400 containing example embodiment fuel assemblies 100 and 200 modified in accordance with example methods. As shown in FIG. 4, four example embodiment assemblies 100 are in controlled locations about a control blade 45a that is anticipated to be used to control the fission chain reaction in the core. According to example methods, assemblies 100 about blade 45a have channels 120 configured in accordance with S210. For example, channels 120 may be thickened, reinforced, and/or fabricated of a material more resistant to deformation than Zircaloy-2. Or, for example, only select sides or walls 120b directly adjacent to control blade 45a may be configured in accordance with S210, including being thickened, reinforced, and/or fabricated of a material more resistant to deformation than Zircaloy-2. Other walls 120a may be unmodified or thinned and/or fabricated of a material equally or less resistant to deformation than Zircaloy-2, in accordance with S220. Example assemblies 200, adjacent to blade 45b that is not to be operated during the fuel cycle or adjacent to no control blade, may be configured in accordance with S220. For example, channels 121 in assemblies 200 may be thinned and/or fabricated of a material equally or less resistant to deformation than Zircaloy-2. Example methods including S100 and S210/S220 may be executed for each assembly to be placed within a core. Alternatively, example methods may be executed only with respect to particular assemblies in order to optimize core operating characteristics. For example, if example fuel assembly channel configuring methods are used in conjunction with other known core configuration methods, the calculated or desired fuel assembly locations and characteristics may require no fuel assembly channel configuring or reconfiguring as in S210 or S220. Example methods may be used as an integral part of core design or as a separate step performed alternatively and/or iteratively with other known methods of core design. For example, a known core design program may output a core map using fuel assembly characteristics with fuel having uniform channel properties. Example methods including S100 and S210/S220 may then be performed on some or all fuel assemblies involved in the map, changing their operational characteristics. The core design program may then be re-executed with the modified fuel assembly characteristics, and this alternating core configuring between example and known methods may continue until no further optimization is possible or desired. Or, example methods may be used as an integral part of otherwise known core design methods, treating reactivity, bow likelihood, and other fuel assembly parameters affected by channel configuring in S210 and S220 as additional variables in the core design process. Example embodiments and methods 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. Variations are not to be regarded as departure from the spirit and scope of the example 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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claims | 1. A nuclear fuel pellet for a nuclear reactor, comprising:a metallic matrix; andceramic fuel particles of a fissile material dispersed in the metallic matrix,wherein the metallic matrix is an alloy of U5-6Zr3-4NbTi wherein a concentration of each element in the metallic matrix is at most 50 molar-%. 2. The nuclear fuel pellet according to claim 1, wherein the concentration of each of the elements in the metallic matrix is at least 5 molar-%. 3. The nuclear fuel pellet according to claim 1, wherein the alloy is a single phase alloy with precipitates constituting less than 5 volume-% of the alloy. 4. The nuclear fuel pellet according to claim 1, wherein the alloy is a High Entrophy Alloy; (HEA). 5. The nuclear fuel pellet according to claim 1, wherein the alloy has a body centered cubic structure. 6. The nuclear fuel pellet according to claim 1, wherein a total concentration of additional elements in the metallic matrix is at most 5 molar-%. 7. The nuclear fuel pellet according to claim 1, wherein the ceramic fuel particles are uniformly dispersed in the metallic matrix. 8. The nuclear fuel pellet according to claim 1, wherein the fissile material is selected from the group consisting of actinide oxide, actinide nitride, actinide silicide and actinide carbide. 9. The nuclear fuel pellet according to claim 1, wherein the fissile material is selected from the group consisting of UO2, U3Si2, U3Si, USi, UN, PuO2, Pu3Si2, Pu3Si, PuSi, PuN, ThO2, Th3Si2, Th3Si, ThSi and ThN. 10. A fuel rod comprising a cladding tube enclosing a plurality of nuclear fuel pellets according to claim 1. 11. A fuel assembly for use in a nuclear reactor, comprising a plurality of fuel rods according to claim 10. |
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claims | 1. A device for converting radiation energy to electrical energy, including:an electrical potential source having a first terminal and a second terminal;a first conductive material electrically coupled to the first terminal;a second conductive material electrically coupled to the second terminal;a third conductive material capacitively coupled to the first conductive material;a fourth conductive material capacitively coupled to the second conductive material; anda radiation receiving area;the third conductive material and fourth conductive material being electrically coupled together to create an electrical current from an electrical potential resulting from radiation received in the radiation receiving area. 2. The device of claim 1, wherein the electrical potential source is a supercapacitor. 3. The device of claim 1, wherein the third conductive material is negatively charged and the fourth conductive material is positively charged. 4. The device of claim 1, wherein the fourth conductive material receives a negative charge from the radiation receiving area and wherein the third conductive material receives a positive charge from the radiation receiving area. 5. The device of claim 1, wherein the first conductive material and the third conductive material are separated by a first electrically isolating material. 6. The device of claim 5, wherein the second conductive material and the fourth conductive material are separated by a second electrically isolating material. 7. The device of claim 1, wherein the electrical current is configured to flow in a pre-selected direction. 8. The device of claim 1, wherein the first terminal comprises a cathode and the second terminal comprises an anode. 9. The device of claim 1, wherein the first terminal comprises a first lead and the second terminal comprises a second lead. 10. The device of claim 9, wherein the first lead and the second lead comprise aluminum. 11. The device of claim 1, wherein the third and fourth conductive materials have an electric potential difference between 100 and 150 volts. 12. The device of claim 1, wherein the third and fourth conductive materials have an electric potential difference between 75 and 100 volts. 13. The device of claim 1, wherein the first conductive material is surrounded by a first oxide material and the second conductive material is surrounded by a second oxide material. 14. The device of claim 13, wherein the first oxide material and the second oxide material comprise aluminum oxide. 15. The device of claim 1, wherein the first, second, third, and fourth conductive materials comprise aluminum. 16. The device of claim 1, wherein the radiation receiving area comprising a noble gas. 17. The device of claim 1, wherein the electrical potential source comprises a battery. 18. The device of claim 1, wherein the first, second, third, and fourth conductive materials are plate shaped. 19. The device of claim 1, wherein first, second, third, and fourth conductive materials each comprises a first plate having a first multitude of teeth and a second plate having a second multitude of teeth, wherein the first multitude of teeth are interlocked with the second multitude of teeth. 20. The device of claim 1, wherein the first, second, third, and fourth conductive materials are cylindrically shaped. 21. The device of claim 1, further comprising a rod positioned in each of the first, second, third, and fourth conductive materials. 22. The device of claim 1, wherein the first, second, third, and the fourth conductive materials are spherically shaped. 23. The device of claim 1, wherein the third and fourth conductive materials have an electric potential difference between about 100 and 1600 volts. 24. The device of claim 1, wherein the third and fourth conductive materials have an electric potential difference between about 100 and 1200 volts. 25. The device of claim 1, wherein the third and fourth conductive materials have an electric potential difference between about 100 and 1000 volts. 26. The device of claim 1, wherein the third and fourth conductive materials have an electric potential difference between about 100 and 800 volts. 27. The device of claim 1, wherein the third and fourth conductive materials have an electric potential difference between about 100 and 400 volts. 28. The device of claim 1, wherein the third and fourth conductive materials have an electric potential difference between about 100 and 200 volts. 29. The device of claim 1, wherein the third and fourth conductive materials have an electric potential difference within a limited proportionality region of a gas in the radiation receiving area. 30. The device of claim 1, further comprising a first transition metal material placed between the third conductive material and the radiation receiving area, and a second transition metal material placed between the fourth conductive material and the radiation receiving area. 31. A device for converting potential energy to electrical energy, including:an electrical potential source having a first terminal and a second terminal;a first conductive material electrically coupled to the first terminal;a second conductive material electrically coupled to the second terminal;a third conductive material coupled to the first conductive material and positioned between the second conductive material and the first conductive material; anda fourth conductive material coupled to the second conductive material and positioned between the first conductive material and the second conductive material;the third conductive material and the fourth conductive material being spaced apart to define a space adapted to receive a gas, andthe third and fourth conductive materials being electrically coupled together to create an electrical flow generated by an electrical potential resulting from a self-ionization of the gas. 32. The device of claim 31, wherein the first, second, third, and fourth conductive materials are cylindrically shaped. 33. A method of generating electrical current, comprising:providing a radiation receiving area for receiving radiation;providing a negatively biased conducive material;providing a positively biased conductive material;causing, by receiving radiation from a radiation source, a plurality of atoms to lose an electron;receiving, by the positively biased conductive material, the plurality of electrons;receiving, by the negatively biased material, a plurality of positively charged particles;the negatively biased conductive material being electrically coupled to the positively biased conductive material to create an electrical current generated by the receiving radiation. 34. The method of claim 33, wherein the radiation receiving area comprising a noble gas. 35. The method of claim 33, wherein the electrical current is configured to flow in a preselected direction. 36. The method of claim 33, wherein the positively biased material and the negatively biased material have a potential difference of 100-150 volts. 37. The method of claim 33, wherein the positively biased material and the negatively biased material have a potential different of 75-100 volts. |
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description | This application claims the benefit of U.S. Provisional Application No. 62/961,964, filed on Jan. 16, 2020. The subject matter thereof is hereby incorporated herein by reference in its entirety. The United States government has rights in this invention pursuant to Contract No. 89233218CNA000001 between the United States Department of Energy and Triad National Security, LLC for the operation of Los Alamos National Laboratory. The present invention generally relates to containment for nuclear materials management. Containers play a pivotal role in nuclear materials management. The complicated nature of constantly crediting containers due to continuously changing requirements, have increased the challenges on defining the containers performance objectives. Issues have emerged that allude to the need for changes in the way containers are managed in specific environments. Factors, such as Material-At-Risk (MAR) and Damage Ratio (DR), are well known and documented for containers used outside the glovebox; however, designing containers for in glove-box applications is more complicated with new requirements. For example, a new requirements documentation (RD) has outlined the primary performance objectives for in-glovebox use including being able to withstand a glovebox fire, a drop or fall from a minimum height of 12 feet and a leak test that is completed by immersing the container in water to a depth of up to 6 inches above the top of the container for a duration of two hours. Accordingly, an improved container for nuclear material management may be beneficial. Certain embodiments of the present invention may provide solutions to the problems and needs in the art that have not yet been fully identified, appreciated, or solved by conventional nuclear material management. For example, some embodiments pertain a container having two different closing mechanism designs including an upright strike-less latch design and a Buttress thread design. In an embodiment, an apparatus for nuclear materials management includes a base container, a container lid, and a rubber-like seal. The apparatus also includes a pair of latches holding the container lid. The pair of latches are tack welded on the container base at or about 120 degrees apart from each other, and a distance between the final placement of the pair of latches and the container lid is dependent on a gasket compression, which is ˜20 percent. Some embodiments pertain a container having two different closing mechanism designs including an upright strike-less latch design and a Buttress thread design. Initial engineering evaluations show both the latch container and threaded design can achieve a DR value of 0.01 after a drop. The threaded design, however, is the solely watertight container after the drop. The container may also be incorporated into other existing container systems, such as the SAVY-4000. Performance Requirement Identification The new RD defines the design requirements for the new general use container designs for glovebox work. The intent behind this document is to capture key aspects of container function and performance considerations across a variety of programmatic process needs for both filtered/vented and hermetic seal containment options. The main drivers for this new container involve criticality requirements for the container to remain watertight under accident scenarios and Material-At-Risk (MAR) credibility for the container to be able to prevent the release of material under accident scenarios. Design Selection The filter utilized in both container types will utilize the new filter that consists of the current ceramic fiber base with the addition of a new perfluorooctyl-trichlorosilane (PFOTS) treatment. The new PFOTS will be hydrophobic at high temperatures will maintaining material integrity under high gamma does rates. FIG. 1 is an image of a strike-less design of a container 100, which includes a base container 102, a container lid 104, and a rubber-like seal 106. Latches 108 to hold container lid 104 are also shown in FIG. 1. Latches in some embodiments are tack welded on container base 102 at or about 120 degrees apart from each other. The distance between the final placement of each latch 108 and lid 104 is dependent on the gasket compression. The desired compression on the gasket is approximately 20%. This is accomplished by estimating the squeeze at 20% compression through the model. Handle 110 on lid 104 is composed of a ⅛″ diameter 316L stainless steel bar, bent in 4 places at 90°. Handle 110 is assembled onto lid 104 by fitting the ends into small stainless steel housings that are tacked welded onto lid 104. The housing for handle 110 is hollow which allows handle 110 to rotate freely along its axis. In some embodiments, such as that shown in FIG. 2, container 200 utilizes a unique buttress threaded design. This unique thread 204 is used to handle extreme axial pulse loading or burst in the axial direction. One side of the thread is perpendicular to the axis while the flank angle is slanted at 45°. The combination equates to a longer thread base for increased shear strength on the threads. In addition, in this embodiment, the features from this thread are relied on for the relief of pressurization. The threads act as a tortuous path for gas when significant pressure builds within the container, resulting in the removal of a filter. The sealing will be compensated by creating a knife edge seal at the contact surface between the lid and the body. This feature assists with keeping the container watertight. In some embodiments, lid 206 is fully machined from an aluminum bronze material. This distinct material was selected so galling between the threads can be mitigated. A new knob was also machined to assist with closing the threaded lid onto the body. Preliminary Results Initial engineering evaluation was conducted on both the threaded and latch designs, testing included water ingress testing and drop testing with a cerium oxide (CeO2) simulant. The initial engineering evaluation provides a basis for design performance where the water ingress was measured before and after drop testing while also measuring the mass loss to estimate the airborne mass loss (g) and respirable mass loss (g). Measuring the mass loss is key to determining the damage ratio (DR value) for each design. Water ingress testing required the use of a test vessel that held the 6″ of water column, timer and a mass balance. See, for example, FIG. 3, which is a diagram illustrating a water ingress testing schematic 300, according to an embodiment of the present invention. Each container was weighed before and after being submersed to determine mass difference. This was done on a Monobloc sr64001 balance with an accuracy of 0.1 grams and a maximum measurement of 64100 grams. The containers were weighted down to eliminate the buoyancy effect that causes the containers. Both containers remained submerged for 2 hours then pulled and effort was made to dry the outer surfaces before post weight measurement were recorded. Table 1 below, shows the pre drop water ingress results assuming a density of water to be 1 g/ml. TABLE 1Pre Drop Water IngressPre-Drop Water Ingress ResultsLatch Design 0.9 gThreaded Design18.4 g The container was dropped using a drop tower and the mass release fraction was measured using a particle counter and a calibrated balance. For actual drop tests, three replicate tests of controlled drops (with the same potential energy of the tested container) were performed to measure the background resuspension aerosol. The background values for the respirable and airborne mass are determined at different potential energy values. These values are variable from drop to drop; and therefore, the background is determined for each drop, especially in the cases of significant powder release from previous experiments. The container potential energy in Joules can be determined from equation (1).PE=m*g*h Equation (1)where m=container mass, g is the gravitational acceleration, and h is the drop height. In each drop, the angle was noted both before the drop and at the time of impact. The angle at impact is determined based on the high speed video footage, using the edge of the impact plate as the zero datum and determining the angle of the container based on that datum using a digital angle finder held against the computer screen. The drop angle chosen was the orientation of center of gravity over top corner of the container (CG over top corner), this orientation is considered to be worst case and will impose the most damage onto the container. The latched design was dropped with a payload weight that is considered to be the maximum for the 3 qt size, the simulant was inclusive with the payload. The container payload was comprised of metal shot to achieve the desired test mass with loose CeO2 powder directly on top of the shot. See, for example, FIG. 4, which is an image 500 illustrating a latch container payload and drop angle determination, according to an embodiment of the present invention. The details of the drop are included in Table 2. TABLE 2Latch Container Drop Testing ResultsCeO2 MAR (g)401.1Gross Weight Pre-Drop (g)9001.5Gross Weight Post-Drop (g)9001.6Released Mass, dm (g)1.8Drop OrientationCG over top cornerPre-Drop Angle46.7°Drop Height (ft)12Drop Energy (joules)324DR Value4.5E−03 FIG. 5 are images 500(a) and 500(b) illustrating latch container impact, according to an embodiment of the present invention. In images 500(a) and 500(b), when the container hit the impact plate, the container deformed and released test powder into the air. Two of the three latches popped open during the impaction, but the lid stayed on. White test powder was spilled on the impact plate, after testing the spilled test powder was swept up and collected, and the remainder was vacuumed up into an analytical open face filter holder. The total collected test powder weight was approximately 1.8037 g. The gross weight of tested container before drop was 9001.5 g, the gross weight after drop was measured as 9001.6 g. FIG. 6 is an image 600 illustrating a latch container post drop impact, according to an embodiment of the present invention. In FIG. 7, CeO2 powder is shown on the impact plate. There is a measurement discrepancy between the initial and final gross weight of the container and the measured released mass. There was a two day time interval between the initial weighing and the drop test. Cerium oxide is hygroscopic, and it is possible the CeO2 powder absorbed water vapor and increased the overall mass of the container. After subtracting the 324 J facility background from the measured respirable mass and airborne mass, the net respirable mass and net airborne release can be found in Table 3. TABLE 3Spilled (or Released) Mass Results of the Latch ContainerRespirable Mass, gAirborne Mass, gBackground at 324 J6.02E−05 +− 5.41E−059.11E−05 +− 8.39E−05 Net Latched Design1.94E−02 +− 1.31E−03 3.22E−02 +− 2.12E−03 The threaded design was dropped with a payload weight that is considered to be the maximum for the 3 qt size, the simulant was inclusive with the payload. FIG. 7 are images 700(a)-(c) illustrating threaded container payload and impact, according to an embodiment of the present invention. In FIG. 7, the container payload was comprised of metal shot to achieve the desired test mass with loose CeO2 powder directly on top of the shot along with the container impact. The details of the drop are included in Table 4. TABLE 4Threaded Container Drop Testing ResultsCeO2 MAR(g)402.5Gross Weight Pre-Drop (g)9003.7Gross Weigh Post-Drop (g)9003.7Released Mass, dm (g)0.0872Drop OrientationCG over top cornerPre-Drop Angle41.5°Drop Height (ft)12Drop Energy (joules)324DR Value2.17E−04 When the container hit the impact plate (see FIG. 7), the container deformed and the lid jumped a thread but stayed on. No visible powder puff was observed in the hi-speed video. However, test powder was observed on the impact plate after the drop. After subtracting the 324 J background from the measured respirable mass and airborne mass, the net respirable mass and the net airborne mass released can be found in Table 5. TABLE 5Spilled (or Released) Mass Results of the Threaded ContainerRespirable Mass, gAirborne Mass, gBackground at 324 J6.02E−05 +− 5.41E−05 9.11E−05 +− 8.39E−05Net Threaded Design5.59E−04 +− 6.81E−058.90E−04 +− 1.10E−04 The latched design container test revealed a vulnerability in a CG over top corner drop. Two latches opened in the latched design drop test, with one of those latches coming completely unlatched. FIG. 8 is an image 800 illustrating latch container post drop deformation, according to an embodiment of the present invention. As shown in image 800, examination of the container after the drop revealed a visible gap between the lid and the container body. The Damage Ratio values for both design types are well below 1% material loss providing a positive closure mechanism with minimum release after an accident scenario and a DR value of at least 0.01. Due to the latch design experiencing enough deformation to create a gap between the lid and body, the container was not tested for water ingress, under the current test requirements the container would have filled entirely with water. In contrast, the threaded design was water resistant after the drop results can be seen in Table 6. TABLE 6Post Water Ingress ResultsPre-Drop Water Ingress ResultsLatch DesignN/AThreaded Design41.1 g The new prototypes show promise for performance testing based on functionality checks and initial engineering evaluations. Functionality checks included opening/closing of each design and the overall engineering judgment on sealing with emphasis on performance while initial engineering evaluations composed of water ingress testing and drop testing with a plutonium simulant for DR, airborne mass and respirable mass values estimations. Water ingress testing was completed under 6″ water column (W.C) for 2 hours while trying to prevent the entry of no more than 50 mil. of water followed by dropping testing at the RRFMC. Glovebox fire testing was not conducted on these containers due to not having access to a convection oven that meets the glovebox fire requirements, this testing is will be completed in the near future. Results show the latched design is watertight at a pristine state while drop testing reveals a 1.8037 g powder loss resulting in DR value of at least 0.01 with minimum airborne mass and respirable mass values, due to the deformation caused at impact the container was not tested and cannot be considered watertight post drop. The threaded design also showed to be water tight at a pristine state while drop testing reveals a 0.00872 g powder loss resulting in a DR value of at least 0.01 with minimum airborne mass and respirable mass values, the threaded design also passed post drop water ingress testing and therefore can be considered water tight after a drop scenario. The threaded design container had about two orders of magnitude less mass loss than the attached design container. Currently, the threaded design container shows more promise than the latched design but, design consideration are being made for better latches. It will be readily understood that the components of various embodiments of the present invention, as generally described and illustrated in the figures herein, may be arranged and designed in a wide variety of different configurations. Thus, the detailed description of the embodiments of the present invention, as represented in the attached figures, is not intended to limit the scope of the invention, but is merely representative of selected embodiments of the invention. The features, structures, or characteristics of the invention described throughout this specification may be combined in any suitable manner in one or more embodiments. For example, reference throughout this specification to “certain embodiments,” “some embodiments,” or similar language means that a particular feature, structure, or characteristic described in connection with the embodiment is included in at least one embodiment of the present invention. Thus, appearances of the phrases “in certain embodiments,” “in some embodiment,” “in other embodiments,” or similar language throughout this specification do not necessarily all refer to the same group of embodiments and the described features, structures, or characteristics may be combined in any suitable manner in one or more embodiments. It should be noted that reference throughout this specification to features, advantages, or similar language does not imply that all of the features and advantages that may be realized with the present invention should be or are in any single embodiment of the invention. Rather, language referring to the features and advantages is understood to mean that a specific feature, advantage, or characteristic described in connection with an embodiment is included in at least one embodiment of the present invention. Thus, discussion of the features and advantages, and similar language, throughout this specification may, but do not necessarily, refer to the same embodiment. Furthermore, the described features, advantages, and characteristics of the invention may be combined in any suitable manner in one or more embodiments. One skilled in the relevant art will recognize that the invention can be practiced without one or more of the specific features or advantages of a particular embodiment. In other instances, additional features and advantages may be recognized in certain embodiments that may not be present in all embodiments of the invention. One having ordinary skill in the art will readily understand that the invention as discussed above may be practiced with steps in a different order, and/or with hardware elements in configurations which are different than those which are disclosed. Therefore, although the invention has been described based upon these preferred embodiments, it would be apparent to those of skill in the art that certain modifications, variations, and alternative constructions would be apparent, while remaining within the spirit and scope of the invention. In order to determine the metes and bounds of the invention, therefore, reference should be made to the appended claims. |
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039379698 | description | Referring to FIG 1, the basic cross-sectional configuration of a collimator in accordance with this invention is shown as corrugated lead foil strips 13 mounted between successive straight strips 12 of lead foil. In FIG. 1 the thickness of the foil is exaggerated, and for 140 KEV gamma rays the preferred thickness would be about 0.010 inches. It should be understood that foils of material other than lead but which are highly opaque to gamma rays could also be used here but lead is preferred from a cost standpoint. The sinusoidal-like corrugations are formed by a technique later to be described, but it should be understood that other forms of regularly corrugated strips such as the triangular form shown in FIG. 9 could also be employed. The corrugated strips are preferably substantially aligned so that triple-thicknesses of metal are avoided, but the employment of straight strips of lead foil between corrugated strips eliminates any severe tolerance problem in forming and assembling the strips. As will later be explained in more detail, the strips may be mounted together with an adhesive such as epoxy and the use of straight strips thus also eliminates the need for more sophisticated fastening techniques, such as welding. As is evident, the open areas formed by the walls of the straight and corrugated strips are the collimating apertures through which gamma rays may pass unhindered, and as previously mentioned the fact that the apertures are not axially symmetric does not seriously affect the performance of the collimator. As shown in FIG. 2, corrugated strips of lead foil may be constructed by passing a strip 50 of lead foil through a pair of substantially meshed gear-like members, such as pinion wire rollers 42, 43. Lead foil strip 50 has a straight portion 51 entering the rollers and a corrugated portion 52 coming out of the rollers. Rollers 42 and 43 are mounted on a support structure 40 having vertical members 41 on which the rollers are rotatably mounted. One of the rollers is turned by a hand crank 44 with a handle 45. The rotation could also be mechanized. Rollers 42 and 43 must be mounted in slightly less than a completely meshed relation to accommodate the thickness of lead foil so that the foil will not be stretched to a lesser thickness at certain points as it is pulled through the rollers. "24 pitch" pinion wire rollers produce corrugated strips which have an effective aperture size of about 1.6 to 1.7 millimeters which lead foil of 0.010 inch thickness. FIG. 3 shows a collimator being assembled from the corrugated and straight lead strips (the size of the corrugations is exaggerated for pictorial clarity). A fixture 30 is useful in assembling the strips since the walls of support sections 32 and 33 mounted on base 31 intersect at a 90.degree. angle and can be used to keep the strips at a 90.degree. angle to the resulting face of the collimator. Using the fixture 30, a collimator may be constructed by the following technique. A flat strip with epoxy spread on one side is placed on support 33, epoxy side up (an epoxy such as "Bondmaster 773B" from PPG Industries may be used). Then, a corrugated strip is placed on top of the epoxied strip. A second flat strip with epoxy spread on both sides is placed on top of the corrugated strip, and a second corrugated strip is then placed upon the flat strip. Preferably the second corrugated strip is aligned with the first one as closely as the eye can tell. This laying down of a flat strip with epoxy on both sides followed by a corrugated strip is continued until the desired size of collimator is achieved. Of course, the lengths of the strips must also be selected in accordance with the desired size of collimator. It is preferable to "load" the collimator structure after every inch or so of strips is built up to insure that the strips are tightly packed together. This can be done by resting a solid lead block on the strips for a short time. In FIG. 3 a partial collimator assembly 34 is shown on the fixture 30. When the desired number of strips have been assembled into a square array of appropriate size, a pair of thin aluminum sheets 22 and 23 may be epoxied onto the square honeycomb-like array 21 as shown in FIG. 4 to build up the assembly 20. At least one such sheet should be used to provide additional structural rigidity to the assembly as well as to protect the edges of the strips from performance-damaging deformations. Two sheets are preferred to provide even more rigidity and protection. The thin aluminum sheets are essentially transparent to gamma rays of 140 KEV energy. After the epoxy in assembly 20 has set, the corners 11 thereof may be removed with a band saw, leaving a circular central portion 10 as the resulting collimator. This circular collimator disc usually serves as an insert in a collimator mounting frame which attaches to the head of a scintillation camera. A schematic side view of the elements in an Anger-type scintillation camera are shown in FIG. 6. Collimator 10 is interposed between a subject 90 containing a radioisotope distribution and a thin sodium iodide crystal 60. A light pipe 70 transmits light from crystal 60 to a matrix of photomultiplier tubes 80. Parallel channel collimator 10 functions essentially to allow only those gamma rays which are emitted from a location in object 90 in a direction substantially perpendicular to crystal 60, i.e., substantially parallel to apertures in collimator 10, to reach crystal 60. Thus a gamma ray 92 from a location 91 interacts in crystal 60 to produce a scintillation 93. A gamma ray headed off at a different angle such as the ray designated 94 will probably be absorbed in one or more walls of collimator 10. All of the functional details of a scintillation camera are well known and need not be further explained here. FIG. 9 shows an alternate form of a corrugated collimator according to this invention. Here corrugated strips 113 have a triangular configuration. A collimator constructed with this form of corrugated strip has a somewhat improved performance which is more reproducible in volume production. Highly regular triangular corrugations are produced by first forming initial corrugated strips in gear-like members 42 and 43 in FIG. 2, and then pressing each strip between male and female forming dies 100, 110 shown in FIGS. 7 and 8. Gear-like members 42 and 43 are either pinion-wire rollers, as previously disclosed, or rollers with triangular teeth. It is vital to form initial corrugated strips in order to have sufficient metal between forming dies 100, 110 such that no tearing or excessive stretching of the lead strips is produced. Thin lead strips are not sufficiently ductile to enable the corrugated strips to be formed directly from straight strips in forming dies 100 and 110. The method of forming highly regular, triangular corrugations thus involves an additional step, and the collimator produced with triangular corrugations is somewhat superior to one with sinusoidal-like corrugations. The corrugation method of collimator construction is also highly advantageously employed to produce converging or diverging channel collimators. The equipment shown in FIGS. 10-14 is employed to form corrugated strips of lead foil which have the characteristics that the triangular corrugations focus to a common point and are tapered, with wider and deeper corrugations at portions remote from the focal point. Because of the tapered corrugations, the strips stack in such a way that they all point to a common focal line. The resultant collimator is a highly satisfactory diverging or converging collimator for use in imaging low energy radiopharmaceuticals. FIG. 10 shows apparatus 120 for producing an initial corrugated strip for a converging or diverging collimator. A base 121 supports in a rotatable fashion a tapered gear-like member 124 on a pair of posts 126, 127. A support arm 122 is mounted on base 121 by way of a pivot structure 123. Support arm 122 carries a second tapered gear-like member 125 on a pair of posts 128, 129. 130 refers to a crank for turning tapered gear-like member 125. 140 refers generally to a locking apparatus for maintaining base 121 and arm 122 in a set relationship when the gear-like members are engaged. When arm 122 is pivoted downward, surfaces 128A and 129A on posts 128 and 129 rest on surfaces 126A and 127A on posts 126 and 127. These surfaces are machined or shimmed such that, when in contact, the degree of meshing or engagement of gear-like members 124 and 125 is proper for receiving and forming an initial corrugated strip. FIGS. 13 and 14 show that gear-like member 124 has tapered triangular teeth to form tapered triangular corrugations in a flat strip. The diameter of gear-like member 124 and the size and taper of triangular teeth thereon determine the focal point of the initial corrugated strip. Preferred method steps for producing an initial corrugated strip will now be detailed in conjunction with FIGS. 10 and 13-15. The first step is to form a flat strip of lead foil with curved top and bottom edges (strip 161A, FIG. 15). This is formed by cutting or stamping curved strip 161 from rectangular piece 160. The next step is to form in strip 161A an initial central corrugation having a tapered triangular configuration. The result is strip 161B with initial corrugation 162. Initial corrugation 162 must be carefully formed to have its axis perpendicular to an intersecting tangent line of the curved strip. A tapered file of the proper size or a specially made tapered forming tool is used to produce the initial corrugation, which is then marked for later reference with paint or ink. Strip 161B is then inserted between gear-like members 124 and 125 with initial corrugation 162 over one of the tapered teeth, and gear-like members 124 and 125 are brought together and locked by locking apparatus 140 into a substantially meshed relationship. Crank 130 is then turned in one direction to form the first half of an initial strip. Preferably, the strip is removed and repositioned with the central reference corrugation between substantially meshed teeth, and crank 130 is turned in the opposite direction to form the second half of an initial corrugated strip. These steps result in a strip such as 161C in FIG. 15. It should be noted that, when strip 161B is inserted in the apparatus 120 of FIG. 10, it is placed on the left end of gear-like members 124 and 125; and as crank 130 is turned, the strip is pulled toward the right. Strip 161C therefore has generally triangular, tapered corrugations which vary in width and depth from center to edge and have rounded corners. The next step is to press initial corrugated strip 161C between male and female forming dies, the male die 150 being shown in FIGS. 11 and 12. This step produces the final corrugated strip 161F which is trimmed to a rectangle while in the forming dies. A large number of these final corrugated strips are made for assembly into a final collimator product. FIG. 16 shows a collimator assembly fixture 170 which is employed in assembling the final product. A base 171 holds a collimator base plate 180 on which individual corrugated and straight strips will be mounted and fastened during the process of gluing individual strips together. Support members 172 and 173 are rigidly mounted to base 171, and alignment arm 176 is pivotally mounted to support member 172 by a pivoting mechanism 174. Alignment arm 176 has a telescoping portion 175 with an alignment wiper arm 177 at the end thereof. A stop block 178 rests on top of base plate 180 during the construction of the first half of the collimator. Final assembly of the collimator follows this procedure. Base 171 is tilted somewhat so that an initial corrugated strip may be rested on stop block 178 such that it touches base plate 180 and points along line 179 to the pivot line of alignment arm 176. A first straight strip with both sides glued is rested on the initial corrugated strip and a second corrugated strip is mounted on the first straight strip. Alignment wiper arm 177 is employed to press the strips together to ensure good contact and to ensure that the second corrugated strip points toward the pivot of aligmnet arm 176. It should be noted that the pivot line is set in terms of the taper of the individual corrugated strips so that the strips quite naturally stack to point toward the pivot line. The alignment arm ensures that the strips are tightly pressed together and proper alignment is maintained. Straight and corrugated strips are added, each time using alignment arm 176 until one half of a collimator is built up. The glue on the strips tends to run down between assembled strips and automatically glues the assembled strips to base plate 180. To assemble the second half of the collimator, stop block 178 is removed and base plate 180 is turned around so that strips can be built up in the other direction. The resultant collimator has triangular channels which generally focus to a short line segment, which is entirely adequate for purpose of gamma imaging with a radiation camera. FIG. 17A shows the imaging properties of a diverging collimator 210A which enables the imaging, on a radiation sensitive transducer 206A, of an object 290A which is larger than transducer 260A. It is called a diverging collimator because its collimating apertures diverge in the transducer to object direction. Dashed lines 211A and 212A portray the diverging field of view of the collimator. FIG. 17B shows the imaging properties of a converging collimator 210B which enables magnification imaging, on a radiation transducer 260B, of an object 290B which is smaller than transducer 260B. It is considered a converging collimator because its collimating apertures converge in the transducer to object direction. From the above it should be apparent that the method of this invnetion is capable of producing all three forms of low energy multichannel collimators--parallel, converging, and diverging channel units. The method results in high quality collimators at low manufacturing cost. FIGS. 18 and 19 illustrate an alternate method of forming initial corrugated strips. A pair of forming tools 273 and 274 pivoted at point 275 press portions of lead foil 161E into tapered grooves of male forming die 150. Forming heads 271 and 272 are tapered to match grooves in forming die 150. After making an initial central corrugation, using head 272 to push foil 161E into the groove in the forming die, head 272 holds the initial corrugation in that groove while head 271 pushes foil into an adjacent groove to form the next corrugation. Both heads are then shifted one groove and head 272 holds the last made corrugation while head 271 forms the next adjacent corrugation. This continues until all the corrugations are formed in an initial corrugated strip, and then the female die is mated to the male die and pressure exerted to make the final corrugated strip. Assembly of corrugated strips made by this alternate method into a final collimator would proceed as set forth above. The above description is intended to be exemplary of general concepts of methods of constructing corrugated collimators, and many changes could be made without departing from the scope of the invention as claimed in the following claims. |
claims | 1. A missile shield assembly comprising:a first beam that is horizontally oriented, the first beam defining a first elongate slot along at least one side of the first beam, the first beam further having a removable frame member that provides access to the first elongate slot;a second beam that is parallel to the first beam, the second beam defining a second elongate slot that faces the first elongate slot, the second beam further having a removable frame member that provides access to the second elongate slot;attachment hardware that releasably attaches the removable frame members to the first and second beams; anda plurality of shield plates, each shield plate being sized to slidably engage the first and second elongate slots such that the shield plates are retained by the first and second beams. 2. The missile shield assembly of claim 1, wherein the missile shield further comprises a plurality of horizontal work platforms. 3. The missile shield assembly of claim 1, wherein the shield plates further comprise an upwardly extending handle. 4. The missile shield assembly of claim 1, wherein the shield plates are formed from a suitable radiation shielding material. 5. The missile shield assembly of claim 4, wherein the shield plates are formed from carbon steel. 6. The missile shield assembly of claim 1, further comprising a second plurality of shield plates and a plurality of additional beams similar to the first and second beams, the additional beams having elongate slots and removable frame members that provide access to the elongate slots, the additional beams being oriented parallel to the first and second beams and spaced to slidably retain the second plurality of shield plates. 7. The missile shield assembly of claim 1, wherein the attachment hardware comprises a plurality of bolts. 8. A missile shield assembly comprising:a support structure including at least a first work platform and a second work platform;a sliding frame structure comprising a plurality of parallel slotted beams, each of the slotted beams defining an elongate slot, wherein the slotted beams are disposed between the first work platform and the second work platform; anda plurality of missile shield plates slidably retained by the slotted beams;wherein at least some of the plurality of parallel slotted beams include a removable frame member and attachment hardware releasably attaching the removable frame member to the associated slotted beam, such that removal of the removable frame member defines a gap that provides access to the elongate slot defined by the associated slotted beam, and such that at least one of the plurality of missile shield plates is removable. 9. The missile shield assembly of claim 8, wherein the attachment hardware comprises a plurality of bolts and further, wherein the removable frame member comprises a plate secured to the slotted beam with the plurality of bolts. 10. The missile shield assembly of claim 8, wherein at least some of the plurality of missile shield plates are slidably retained by the slotted beams in a row, such that the missile shield plates in the row are each removable by sliding the missile shield plates to the gap defined by removal of the removable frame member and lifting the missile shield plate away from the slotted beam. 11. The missile shield of claim 10, wherein each of the missile shield plates includes a handle to facilitate lifting the missile shield plate. 12. A missile shield assembly comprising:a support structure means including a plurality of work platforms;a sliding frame structure including a plurality of opposed elongate slots defining a plurality of horizontal tracks, a plurality of removable frame members and attachment hardware that releasably attaches the removable frame members to the sliding frame structure; anda plurality of carbon steel shield plates slidably retained in the plurality of horizontal tracks;wherein removing the removable frame members provides access to the plurality of horizontal tracks such that the plurality of shield plates are individually removable. |
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claims | 1. An extreme ultraviolet light source apparatus for generating extreme ultraviolet light from plasma generated by irradiating a target with a laser beam, said apparatus comprising:a chamber in which the extreme ultraviolet light is generated;a target supply unit configured to supply the target into said chamber;a laser beam focusing mirror arranged in said chamber to focus the laser beam;a collector mirror having a reflection surface composed of a multilayer film and arranged in said chamber to collect and output the extreme ultraviolet light emitted from the plasma;a first shielding member including at least one metal selected from the group consisting of: silicon (Si), zirconium (Zr), molybdenum (Mo), lithium (Li), aluminum (Al), and an alloy including at least one of the metals, and arranged between said target supply unit and the plasma, said first shielding member being arranged apart from the reflection surface of said collector mirror;a second shielding member including at least one metal selected from the group consisting of: silicon (Si), zirconium (Zr), molybdenum (Mo), lithium (Li), aluminum (Al), and an allo including at least one of the metals, and arranged on an inner wall of said chamber; anda third shielding member including at least one metal selected from the group consisting of: silicon (Si), zirconium (Zr), molybdenum (Mo), lithium (Li), aluminum (Al), and an alloy including at least one of the metals, and arranged between said laser beam focusing mirror and the plasma. 2. The extreme ultraviolet light source apparatus according to claim 1, wherein said first shielding member includes a first shield plate formed with an opening through which the target passes. 3. The extreme ultraviolet light source apparatus according to claim 1, wherein:said second shielding member includes a second shield plate configured to cover the inner wall of said chamber; andsaid apparatus further comprises a cooling device arranged on said second shield plate and configured to cool said second shield plate. 4. The extreme ultraviolet light source apparatus according to claim 1, wherein at least one of said first, second, and third shielding members includes a coating. 5. The extreme ultraviolet light source apparatus according to claim 4, wherein a material of said coating is the same as a material of said multilayer film of said collector mirror. 6. The extreme ultraviolet light source apparatus according to claim 1, wherein said at least one of the inner wall of said chamber and said target supply unit includes stainless steel. 7. The extreme ultraviolet light source apparatus according to claim 1, wherein:said target supply unit includes a charge supplying device configured to charge the target, and a deflection device configured to deflect said target charged by said charge supplying device to supply said target to a plasma emission point andsaid first shielding member is arranged between said deflection device and the plasma. 8. The extreme ultraviolet light source apparatus according to claim 7, wherein said charge supplying device includes any one of a discharge electrode, an electron gun, a plasma tube, and a plasma torch. 9. The extreme ultraviolet light source apparatus according to claim 7, wherein said deflection device includes:a first pair of electromagnets oppositely arranged in parallel with each other with a trajectory of the target in between;a second pair of electromagnets orthogonal to said first pair of electromagnets and oppositely arranged in parallel with each other with the trajectory of the target in between; anda power supply configured to supply currents to said first and second pairs of electromagnets such that the same magnetic poles face to each other and the different magnetic poles adjoin each other. 10. The extreme ultraviolet light source apparatus according to claim 1, wherein:said target supply unit includes a charge supplying device configured to charge said target, and an accelerator configured to accelerate said target charged by said charge supplying device to supply said target to a plasma emission point; andsaid first shielding member is arranged between said target accelerator and the plasma. 11. The extreme ultraviolet light source apparatus according to claim 1, further comprising:a sensor arranged outside of said chamber;an observation window for said sensor, said observation window being provided in said chamber; anda fourth shielding member including at least one metal selected from the group consisting of silicon (Si), zirconium (Zr), molybdenum (Mo), lithium (Li), aluminum (Al), and an alloy including at least one of the metals, and configured to shield said observation window from debris emitted from the plasma. 12. The extreme ultraviolet light source apparatus according to claim 1, further comprising:a metal grid arranged in front of the reflection surface of said collector mirror; anda power supply configured to induce positive potential in said metal grid to repulse ion debris. 13. The extreme ultraviolet light source apparatus according to claim 12, wherein said metal grid is formed of material having a higher transmittance for the extreme ultraviolet light than that of said at least one of the inner wall of said chamber and said target supply unit. 14. The extreme ultraviolet light source apparatus according to claim 1, further comprising:electromagnets arranged to surround the plasma, and configured to form a local magnetic field in surroundings of the plasma, and thereby, guide a moving direction of ions emitted from the plasma so as to divert the moving direction of ions from the reflection surface of said collector mirror. 15. The extreme ultraviolet light source apparatus according to claim 14, wherein said electromagnets are arranged in an obscuration area which corresponds to an angle range in which the extreme ultraviolet light collected by said collector mirror is not utilized in an exposure unit. 16. The extreme ultraviolet light source apparatus according to claim 1, wherein said third shielding member includes a third shield plate formed with an opening through which the laser beam passes. |
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description | This patent application claims the priority benefit under 35 U.S.C. § 119(e) of U.S. Provisional Application No. 62/547,389 filed on Aug. 18, 2017, the contents of which are herein incorporated by reference. The present invention is directed to improved methods of determining a core design parameter of a nuclear reactor and further verifying that such core design parameter is within an expected range and if not, taking further actions. As shown schematically in FIGS. 1 and 2, a Nuclear Instrumentation System (NIS) 10 is a safety related system which utilizes three detector types—commonly Source Range 12, Intermediate Range 14 and Power Range 16—to provide adequate monitoring of a nuclear reactor 18 over approximately 11 decades of reactor power. Each NIS detector 12, 14, 16 can be used as an input to a reactivity computer (not shown) to calculate various core design parameters of the nuclear reactor, e.g., see U.S. Pat. No. 4,877,575, the contents of which are incorporated by reference herein. The signal from each detector 12, 14, 16 represents a directly proportional value to neutron flux and is used to solve the Inverse Point Kinetics Equation or other reactivity equations. A reactivity computer can connect directly to a detector 12, 14, 16; however, such connection renders the detector 12, 14, 16 unable to perform its safety related function. Accordingly, it is desirable to use isolated outputs (i.e., outputs from the NIS cabinet 20, 22, 24) when possible so the detector 12, 14, 16 can retain its safety related function. Isolated outputs from the NIS cabinets 20, 22, 24 are typically a 0-5V or 0-10V output based on the range of detector current or pulses. For example, without limitation, output from the Westinghouse Compensated Ion Chamber Intermediate Range Detector ranges from 10−11 to 10−3 amperes and the processing cabinet produces a 0-5V or 0-10V output. The log current amplifier which converts the current or pulse signals to a voltage output signal may typically only have the potential to adjust the calibration of the voltage output at several points in the entire detector span. For instance, one style of the log current amplifier used in the Westinghouse Intermediate Range detector has a span of 8 decades 10−11 to 10−3 amperes range, but can only adjust the output at three points; 10−11, 10−7 and 10−4 amps. While this is adequate for normal plant operation needs, it is not as desirable for use in high precision measurements such as those carried out with a reactivity computer since other decades away from the aforementioned adjustable points in the entire range may still not be within the expected tolerance. Reactivity represents a rate of change of reactor power, so if the rate of change of the detector signal is not calibrated properly, then the corresponding reactivity calculation would be calculated incorrectly. Accordingly, there exists room for improvement in the methods and systems for carrying out such measurements. Embodiments of the present invention utilize as-left calibration information to improve the accuracy of the isolated voltage output from a NIS cabinet to support making the most accurate and precise measurements. As one aspect of the present invention a method of determining a core design parameter of a nuclear reactor is provided. The method comprises: calibrating an isolated voltage output from a NIS cabinet associated with the nuclear reactor using a calibrated signal source as an input to the NIS cabinet; recording values of the calibrated signal source used in the calibrating and corresponding values of the output voltage from the calibrating in an as-left cabinet calibration data table; using a computing device connected to the isolated voltage output from the NIS cabinet, converting the voltage output signal to a converted detector signal using at least some of the values in the as-left cabinet calibration data table in an improved signal conversion equation; and using the computing device, employing the converted detector signal to determine the core design parameter. The method may further comprise comparing the measured core design parameter to a predicted core design parameter to determine if the measured core design parameter is within an acceptable limit. The method may further comprise determining from the comparing that the measured core design parameter is not within the acceptable limit of the predicted core design parameter and taking a further action in response thereto. The calibrated signal source may comprises a calibrated current source. The converted detector signal may comprise a converted detector current signal. Solving an equation using the converted detector signal may comprise solving the Inverse Point Kinetics Equation using the converted detector current signal and one or more additional nuclear design constants as inputs. Determining a core design parameter may comprise determining at least one of Isothermal Temperature Coefficient, Boron Endpoint or Control Rod Worth. The calibrated signal source may comprise a calibrated current pulse source. The converted detector signal may comprise a converted detector pulse signal. The computing device may comprise a reactivity computer. The isolated voltage output from the NIS cabinet may correspond to an input received by the NIS cabinet from a source range detector monitoring the nuclear reactor. The isolated voltage output from the NIS cabinet may correspond to an input received by the NIS cabinet from an intermediate range detector monitoring the nuclear reactor. The isolated voltage output from the NIS cabinet may correspond to an input received by the NIS cabinet from a power range detector monitoring the nuclear reactor. As another aspect of the present invention, a computing device is adapted to: receive an isolated voltage output from an NIS cabinet; convert the isolated voltage output to a converted detector signal using at least some values from an as-left calibration data table created from a previously carried out calibration of the isolated voltage output using a calibrated signal source as an input to the NIS cabinet; and determine a core design parameter using the converted detector signal. The computing device may be further adapted to compare the core design parameter to an expected core design parameter. The present invention will now be described more fully hereinafter with reference to the accompanying drawings, in which examples of the invention are shown. The invention may, however, be embodied in many different forms and should not be construed as limited to the examples set forth herein. Rather, these examples are provided so that this disclosure will be thorough and complete, and will fully convey the scope of the invention to those skilled in the art. Like numbers refer to like elements throughout. FIG. 3 is a flowchart showing the general steps of a method 100 for determining a core design parameter of a nuclear reactor in accordance with an example embodiment of the present invention. Referring to FIG. 3, as well as to FIG. 1, method 100 begins by calibrating an isolated voltage output from one of the NIS cabinets 20, 22, 24 associated with the nuclear reactor 18 such as shown at 110. Such calibration is carried out by providing a range of inputs from a calibrated signal source (e.g., a calibrated current source, a calibrated pulse generator) as inputs to the NIS cabinet 20, 22, 24 and measuring the corresponding voltage output by the NIS cabinet 20, 22, 24 resulting from each calibrated input of the range of inputs. Next, or generally concurrently with step 110, the values of the calibrated signal source used in the calibrating of step 110 and the corresponding values of the output voltage are recorded in an as-left cabinet calibration data table, as shown at 120. Such table may be in the form of a physical (i.e., hardcopy) table, an electronic table accessible via software, or any other suitable form. As shown at 130, at least some of such values from the previously determined as-left cabinet calibration data table are utilized by a computing device (e.g., without limitation, a reactivity computer or other suitable computing device) in an improved signal conversion equation which is used to convert the isolated voltage output from the NIS cabinet 20, 22, 24 to a converted detector signal. The converted detector signal is then used to determine a core design parameter of the nuclear reactor, such as shown at 140. As shown at 150, such determined core design parameter may be compared to an expected core design parameter in order to determine if the nuclear reactor 18 is operating as expected or if a potential problem and/or safety concern exists. If such a deviation exists, subsequent testing and/or precautionary measures may be carried out with respect to the reactor 18, such as generally shown at 160. Having thus described a general method of carrying out concepts of the present invention, an example of a more particular method directed to output from an intermediate range detector 14 will now be discussed. For Westinghouse Compensated Ion Chamber Intermediate Range detectors, such as the IR detector 14 shown in FIG. 1, the isolated voltage output signals leaving the NIS cabinet 22 are in the range of 0-5 VDC, which correspond to IR detector flux measurements of 10−11-10−3 amperes. The standard flux signal current conversion which is presently carried out is represented by the following equation:I(V)=α1×10(α2V+α3)+α4 (1) where: I(V) is the detector flux value in amperes corresponding to IR channel output voltage, V α1 is the applied signal gain (typically=1) α2 is the ratio of the detector flux decades spanned to the output voltage span (typically=8/5) α3 is the minimum detector flux decade (typically=−11) α4 is the value of the current bias present when the IR channel output voltage is equal to zero (typically=0) Such standard flux signal current conversion provides an acceptable representation of detector flux, however, it assumes an ideal (i.e., perfect) calibration/alignment of the NIS drawer isolated voltage output, which is not typical in reality. Furthermore, the Dynamic Rod Worth Measurement (DRWM) technique, such as described by Y. A. Chao, D. M. Chapman, D. J. Hill, L. R. Grobmyer, “Dynamic Rod Worth Measurement,” Nuclear Technology Volume 132 Number 3, December 2000, p403-412, the contents of which are incorporated by reference herein, requires high accuracy flux measurements over 3-4 decades of neutron flux compared to single decade flux measurements utilized for traditional bank worth measurement techniques. In order to achieve the highest level of measurement accuracy for IR detector flux, embodiments of the present invention utilize the actual representation of the calibration/alignment of the NIS drawer isolated voltage output condition when carrying out the flux signal current conversion. FIG. 4 is a schematic diagram showing components and signals of an example Intermediate Range Detection portion 50 of a Nuclear Instrumentation System 10 in accordance with an example embodiment of the present invention. FIG. 5 is a flowchart showing various steps of a method 200 in accordance with an example embodiment of the present invention. Referring to FIG. 5, method 200 begins at step 210 by calibrating an isolated voltage output from an NIS cabinet, i.e., intermediate range cabinet 22, associated with the nuclear reactor 18. Such calibrating is carried out by a technician using a calibrated current source. As shown in step 220, during such calibration, the input currents used and resulting output voltages are recorded in an as-left cabinet calibration data table, such as the example shown in Table 1 below. TABLE 1A sample data table to document IsolationAmplifier alignment checksInjected currentMeasured As leftNominal(amps)(VDC)(VDC)00.000 10−110.188 10−100.65110−91.25310−81.87510−72.50010−63.12510−53.75010−44.37510−35.000 Next, as shown in step 230, the isolated voltage output from the NIS cabinet 22 is connected to a reactivity computer, such as a laptop or other suitable computing device. Next, in step 240, an improved signal conversion equation is utilized by the aforementioned computing device to convert the voltage output signal to a custom (i.e., NIS drawer-specific) flux signal current conversion which incorporates the calibration/alignment information of the particular NIS cabinet drawer. More particularly, the computing device utilizes the following improved signal conversion equation: I ( V ) = 10 { [ ( V - V low ) ( V high - V low ) × lo g ( I high I low ) ] + lo g ( I low ) } ( 2 ) where: I(V) is the detector flux value in amperes corresponding to IR channel output voltage, V Vlow is the lower bounding as-left voltage for the measured IR channel output voltage interval Vhigh is the upper bounding as-left voltage for the measured IR channel output voltage interval Ilow is the lower bounding detector current corresponding to Vlow i.e., injected current+idling current (typically 10−11 amperes) Ihigh is the upper bounding detector current corresponding to Vhigh i.e., injected current+idling current (typically 10−11 amperes) As a broad check of such equation, consider the case of the standard flux signal current conversion, where the piecewise equation is reduced to one large interval over the entire detector range: I ( V ) = 10 { [ ( V - 0 ) ( 5 - 0 ) × lo g ( 10 - 3 10 - 11 ) ] + lo g ( 10 - 11 ) } where: I(V) is the detector flux value in amperes corresponding to IR channel output voltage, V Vlow=0 VDC Vhigh=5 VDC Ilow=10−11 amperes Ihigh=10−3 amperes Substituting the above bounding values results in the following equation, which is equivalent to the standard flux signal current conversion with default gain and bias/offset coefficients. I ( V ) = 10 [ ( V 5 × 8 ) - 11 ] A practical demonstration of the custom flux signal current conversion is shown in FIG. 6. During a recent plant startup, Intermediate Range detector isolated voltage output data was measured and the corresponding calibration/alignment information was obtained. The measured voltage data was converted to detector flux data via both the standard (i.e., equation 1 above) and custom (i.e., equation 2 above) flux signal current conversions. The converted flux signal data was then processed with the standard equations used by Westinghouse reactivity computers to calculate core reactivity (i.e., Inverse Point Kinetics Equations with Stiffness Confinement Method). FIG. 6 provides a graphical comparison of the flux signal data converted from the isolated voltage outputs between the standard and custom flux signal current conversions during a portion of the plant startup physics testing. Next, in step 250, the converted detector current signal determined in step 240 is used by the reactivity computer, along with one or more additional nuclear design constant as inputs, to solve the Inverse Point Kinetics Equation. Next, in step 260, the reactivity computer is used in determining at least one core design parameter, such as Isothermal Temperature Coefficient, Boron Endpoint or Control Rod Worth. These are calculations that are done in the computer. In some cases, additional plant signals or input are used, such as the temperature coefficient uses a signal for moderator temperature in the calculation with reactivity. The boron endpoint calculation uses the measured reactivity and value of what the boron concentration inside the RCS is as determined by a titration sample done in the chemistry lab. Next, in step 270, the determined core design parameter is then compared to a predicted parameter to determine if the measured core design parameter is within acceptable limits. If such measured parameter is not within limits, then additional testing, analysis or restrictions may be required. Additional follow up steps for any failure would be to initially determine if there were any measurement process errors, correct the error and re-measure or re-analyze the core parameter. If the control rod worth measurement is not within limits and no measurement process error is identified, then the control rods would be measured by an alternate means, such as by using the boration/dilution method or bank exchange method (both methods identified in ANSI/ANS-19.6.1 standard). From the foregoing description it is thus to be appreciated that the present invention provides an improved method for carrying out reactivity measurements/determinations for a nuclear reactor. 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 arrangements disclosed are meant to be illustrative only and not limiting as to the scope of invention which is to be given the full breadth of the claims appended and any and all equivalents thereof. |
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description | The present application is based on, and claims priority from, Korean Application NO. 2004-35571, filed on May 19, 2004, the disclosure of which is incorporated by reference herein in its entirety. 1. Field of the Invention The present invention relates to a method for fabricating a sintered duplex nuclear fuel pellet, and more particularly to a method for fabricating a sintered duplex nuclear fuel pellet consisting of a cylindrical inner portion composed of UO2—Gd2O3 and an annular outer portion composed of UO2—Er2O3 or UO2, which is capable of preventing the occurrence of cracks due to the difference in the densification rate between the two portions upon sintering. 2. Description of the Related Art Generally, the most widely used sintered pellet as a nuclear fuel is a sintered uranium dioxide (UO2) pellet. The sintered uranium dioxide pellet mainly has a height of about 10 mm and a diameter of about 8 mm in a cylindrical shape for practical application. The sintered UO2 pellet contains 1-5% by weight of U235 and produces nuclear fission energy while U235 decays by the action of neutrons during use in a nuclear reactor. In addition to the sintered UO2 pellet, a sintered burnable absorber pellet containing neutron absorbing materials, such as gadolinium (Gd) or erbium (Er), is used in a nuclear reactor to control the numbers of neutrons. The sintered burnable absorber pellet has generally the same size as the sintered UO2 pellet. Gd or Er is uniformly distributed throughout the whole sintered UO2 pellet. The sintered burnable absorber pellet is commonly represented by (U,Gd)O2 or (U,Er)O2, and is fabricated in accordance with the following procedure. First, UO2 powder is mixed with Gd2O3 or Er2O3 powder, and pulverized to prepare a UO2—Gd2O3 powder or UO2—Er2O3 powder. At this time, the Gd2O3 content and the Er2O3 content are limited to about 15% or less and 4% by weight or less, respectively. Thereafter, the pulverized powder is subjected to compression molding to produce a compact, which is sintered by heating under a reducing gas atmosphere at 1,600-1,800° C. for 2-4 hours. The density of the compact is about 50-70% of the theoretical value, and that of the sintered pellet is about 95% of the theoretical value. During the sintering, Gd or Er is dissolved into the lattice structure of UO2. The reducing gas may be hydrogen gas, or a mixed gas of hydrogen gas with at least one gas selected from water vapor, inert gases and carbon dioxide. Korean Patent No. 0281169 (published on Nov. 15, 2000), issued to Korea Atomic Energy Research Institute, describes that a duplex structure consisting of a cylindrical inner portion and an annular outer portion composed of different materials is advantageous over a structure using only a sintered (U,Gd)O2 or (U,Er)O2 pellet as a sintered burnable absorber pellet, because the duplex structure exhibits improved nuclear performance in a nuclear reactor. FIG. 1 is a schematic view showing the structure of a conventional sintered duplex nuclear fuel pellet 10 consisting of a cylindrical inner portion 1 and an annular outer portion 2. The sintered duplex burnable absorber nuclear fuel pellet is fabricated by charging different nuclear fuel powders into the inner and outer portions, molding the portions to produce a duplex compact, and sintering the duplex compact. However, the sintering process may have a number of problems based on the fact that the inner portion and the outer portion are composed of different materials. That is, the annular outer portion is composed of a mixed powder UO2—Er2O3 and the cylindrical inner portion is composed of a mixed powder UO2—Gd2O3. During the sintering process (also, referred to “densification”) for increasing the density of the compact while contracting, a large difference in the densification rate of the cylindrical inner portion and the annular outer portion composed of different materials is caused, which generates an undesirable internal stress at the interface between both portions. Consequently, serious interstices or cracks occur at the interface of the final sintered duplex nuclear fuel pellet. In an effort to overcome these problems, U.S. Pat. No. 4,678,629 suggests a method comprising sintering a cylindrical inner portion (UO2—Gd2O3) and an annular outer portion (UO2) separated from each other, and fitting the sintered cylindrical inner portion into the sintered annular outer portion. However, since this method requires precise fabrication and processing of the sintered portions, it is not an effective approach for practical application. Alternatively, M. Fisher (J. Nucl. Mater., 138, 242-247 (1986)) reported a method for fabricating a sintered duplex pellet consisting of a ThO2 outer portion and a UO2 inner portion through two-step sintering. According to this method, the UO2 inner portion is first sintered, fitted into a ThO2 compact, and further sintered to fabricate a sintered pellet having no cracks. However, this method further involves fitting of the sintered UO2 inner portion into the ThO2 compact, which renders the overall processes more complicated. In addition, since this method does not suggest a basic solution to a large difference in the densification rate between the cylindrical inner portion and the annular outer portion, the occurrence of cracks still cannot be avoided, resulting in deterioration of the strength of the sintered pellet and damage during handling. In particular, the thermal conductivity is reduced, causing a reduction in the operational performance in a nuclear reactor. Thus, there is a need in the art for a method for fabricating a sintered duplex burnable absorber nuclear fuel pellet free from cracks derived from the difference in the densification rate between a cylindrical inner portion and an annular outer portion composed of different materials upon sintering. The present invention has been made in view of the above problems of the prior arts, and it is an object of the present invention to provide a method for fabricating a sintered duplex nuclear fuel pellet consisting of an cylindrical inner portion composed of UO2—Gd2O3 and an annular outer portion composed of UO2—Er2O3 or UO2, which is capable of preventing the occurrence of cracks at the interface between the two portions by decreasing the difference in the densification rate of the two portions upon sintering. In order to accomplish the above object of the present invention, there is provided a method for fabricating a sintered duplex nuclear fuel pellet, comprising the steps of: preparing a first powder composed of a material selected from the group consisting of UO2 and UO2—Er2O3, and a second powder composed of UO2—Gd2O3 and a sintering additive; producing a duplex compact consisting of an annular outer portion composed of the first powder and a cylindrical inner portion composed of the second powder; and sintering the duplex compact under a reducing gas atmosphere, wherein the sintering additive contains manganese of 0.001% to 2% by weight based on the total weight of the cylindrical inner portion(or the second powder). Preferably, the Er2O3 content in the mixture UO2—Er2O3 constituting the first powder may be about 4% by weight or less, and the Gd2O3 content in the mixture UO2—Gd2O3 constituting the second powder may be in the range of about 4% to about 15% by weight. Preferably, the density of the sintered duplex fuel nuclear pellet may be more than 94% of theoretical value. Preferably, the sintering additive may be at least one selected from the group consisting of pure manganese, manganese oxide (MnO), manganese sulfide (MnS), manganese fluoride and manganese chloride. It is more preferable to use manganese compounds, such as manganese oxide (MnO) and manganese sulfide (MnS), having a relatively high melting point(e.g. 1785° C. and 1610° C., respectively). The reducing gas may be hydrogen gas, or a mixed gas of hydrogen gas with at least one gas selected from water vapor, inert gases and carbon dioxide. The duplex compact can be sintered in the temperature range of between about 1,600° C. and about 1,800° C. In one embodiment of the present invention, at least a part of the UO2 may be replaced by PuO2. The present invention is characterized by the addition of the sintering additive containing about 0.001% to about 2% by weight of manganese, such as pure manganese or the manganese compound, based on the total weight of the cylindrical inner portion, to the cylindrical inner portion. Manganese would be dissolved into the lattice of UO2 and Gd2O3 during sintering and then generate the lattice defects, such as uranium vacancies, because of the difference in valency. Thus, the pure manganese or the manganese compound added promotes the densification of the UO2—Gd2O3 constituting the cylindrical inner portion to reduce the generation of an internal stress between the cylindrical inner portion and the annular outer portion. Generally, the UO2—Er2O3 constituting the annular outer portion shows a similar densification behavior to that of pure UO2, but the UO2—Gd2O3 constituting the cylindrical inner portion has a very low densification rate in a predetermined sintering temperature range. For these reasons, prior arts have a problem that the occurrence of cracks may be induced due to an unwanted stress generated at the interface of a sintered pellet. To solve this problem, the present invention suggests a method for decreasing a large difference in the densification rate by adding pure manganese or the manganese compound as a sintering additive to the powder constituting the cylindrical inner portion. The manganese used herein is dissolved into the lattice structure of UO2 to promote the sintering of the duplex compact, and to increase the densification rate of the UO2—Gd2O3 constituting the cylindrical inner portion, thereby decreasing the difference in the densification rate between the UO2—Gd2O3 constituting the cylindrical inner portion and the UO2—Er2O3 or pure UO2 constituting the annular outer portion. Regardless of the kinds of the manganese compound, i.e., manganese oxide (MnO), manganese sulfide (MnS), manganese fluoride and manganese chloride, the desired sintering promotion effects are attainable. In addition, the method of the present invention can be applied to the fabrication of another sintered duplex nuclear fuel pellet wherein part of the UO2 is replaced by plutonium dioxide (PuO2), which is a nuclear material having the same lattice structure as UO2. Hereinafter, the principle and functions of the present invention will be explained in more detail with reference to the accompanying drawings. FIG. 2 is a graph illustrating the effect of manganese oxide (MnO) as a sintering additive, and shows changes in the densification rate of respective compounds which can be used to fabricate a sintered nuclear fuel pellet, with increasing temperatures. Referring to FIG. 2, UO2-2 wt % Er2O3 (a) constituting an annular outer portion has the highest densification rate at 1,220° C., and UO2-10 wt % Gd2O3 (b) constituting a cylindrical inner portion has the highest densification rate at 1,510° C. Accordingly, the two portions undergo a large difference in the densification rate during the sintering. That is, the cylindrical inner portion is densified at a higher temperature than the annular outer portion. However, when the manganese oxide (MnO) is added in amounts of 0.01%, 0.02%, 0.05% and 0.1% by weight to UO2-10 wt % Gd2O3 (denoted by c, d, e and f in FIG. 2, respectively), based on the total weight of the inner portion, the densification behavior of UO2-10 wt % Gd2O3 changes to be similar to that of the UO2-2 wt % Er2O3 (denoted by “a”). Specifically, as the amount of the manganese oxide (MnO) added increases from 0.01% to 0.1% by weight, temperatures showing the highest densification rate are close to 1,220° C. As demonstrated experimentally above, the densification of the compact is drastically increased by the addition of the manganese compound to the mixed powder UO2—Gd2O3. These results represent that manganese would be dissolved into the lattice of UO2 and Gd2O3 at the early stage of sintering and promote the densification by generating the lattice defects, such as uranium vacancies. In addition, it could be confirmed that the difference in the densification rate between the inner portion and the outer portion decreases with increasing sintering temperatures. In conclusion, a sintered duplex pellet having no cracks at the interface between the inner and outer portions can be fabricated by adding a sintering additive containing manganese to the mixed powder UO2—Gd2O3 constituting the inner portion, producing a compact, and sintering the compact. The sintering additive contains manganese of 0.001% of 2% by weight based on the total weight of the inner portion. When the amount of manganese added is less than 0.001% by weight, sufficient sintering promotion effects are not attainable. On the other hand, when the amount exceeds 2% by weight, characteristics inherent to the mixed powder UO2—Gd2O3 constituting the inner portion may be degraded. It is preferable that the sintering additive may be added as pure manganese or manganese compound. The present invention will now be described in more detail with reference to the following specific examples. These examples illustrate a method for fabricating a sintered duplex nuclear fuel pellet consisting of an inner portion composed of (U,Gd)O2 and an outer portion composed of (U,Er)O2. First, a mixture of 2% by weight of Er2O3 powder and UO2 powder was charged into a tubular mixer and was then mixed for 1 hour to prepare a mixed powder UO2-2 wt % Er2O3 for an annular outer portion of a duplex nuclear fuel. Separately, a mixture of manganese oxide (MnO) and Gd2O3 powder was subjected to ball milling using zirconia balls for 12 hours to prepare a MnO-containing Gd2O3 powder. Then, the powder thus prepared was mixed with UO2 powder in a tubular mixer for 1 hour, and was then pulverized in a pestle and mortar for 10 minutes to prepare a mixed powder UO2-10 wt % Gd2O3-0.1 wt % MnO for a cylindrical inner portion. The mixed powder UO2-2 wt % Er2O3 was charged into the annular outer portion and the UO2-10 wt % Gd2O3-0.1 wt % MnO was charged into the cylindrical inner portion, which was then subjected to compression molding to produce a duplex compact. The duplex compact is shown in FIG. 1. A number of processes for duplex compaction are already known. In this example, the compaction was performed in accordance with the process described in Korean Patent No. 0354544 (published on Sep. 16, 2002), issued to Korea Atomic Energy Research Institute, et al., which is incorporated herein by reference. Thereafter, the duplex compact was sintered as follows. The compact was heated to 1,700° C. at a rate of 5K/min., and was then maintained under a reducing gas atmosphere at 1,700° C. for 4 hours to fabricate a sintered nuclear fuel pellet. At this time, hydrogen gas, or a mixed gas of hydrogen gas with at least one gas selected from water vapor, inert gases and carbon dioxide, may be used as the reducing gas. In this example, a mixed gas H2-3% CO2 was used. The density of the sintered pellet was about 97% of the theoretical density. The sintered nuclear fuel pellet thus fabricated was polished along a face parallel to an axial direction. The occurrence of cracks and defects at the interface was observed. FIG. 3a is a photograph of the overall cross-section the sintered duplex nuclear fuel pellet fabricated in Example 1. FIG. 3b is an optical microscope image showing the vicinity of the interface of the sintered duplex nuclear fuel pellet. As shown in FIGS. 3a and 3b, the addition of manganese oxide (MnO) to the inner portion makes it possible to fabricate the sintered duplex nuclear fuel pellet whose interface is clean without the occurrence of cracks. As can be seen from the graph shown in FIG. 2, the inner portion composed of the mixed powder UO2-10 wt % Gd2O3 containing 0.1% by weight of manganese oxide (MnO), corresponding to 0.077% by weight of manganese, has the maximum densification rate at 1,280° C., which is similar to the temperature (about 1,220° C.) at which the outer portion composed of the mixed powder UO2-2 wt % Er2O3 has the maximum densification rate. Accordingly, the final sintered duplex nuclear fuel pellet can be fabricated without the occurrence of interfacial cracks. By a similar procedure to that in Example 1, a mixture of 2% by weight of Er2O3 powder and UO2 powder was charged into a tubular mixer and was then mixed for 1 hour to prepare a mixed powder UO2-2 wt % Er2O3 for an annular outer portion of a duplex nuclear fuel. Separately, 10% by weight of Gd2O3 powder, which had been previously subjected to ball milling using zirconia balls for 12 hours, was added to UO2 powder, mixed in a tubular mixer for 1 hour, and pulverized in a pestle and mortar for 10 minutes to prepare a mixed powder UO2-10 wt % Gd2O3 for an inner portion, whose composition was different from that prepared in Example 1. That is, the procedure was performed in the same manner as in Example 1, except that manganese oxide was not added to the powder constituting the inner portion. The mixed powder UO2-2 wt % Er2O3 was charged into the outer portion and the mixed powder UO2-10 wt % Gd2O3 was charged into the inner portion. Thereafter, the two portions were compacted, and then sintered to fabricate a sintered nuclear fuel pellet. The sintered nuclear fuel pellet thus fabricated was polished along a face parallel to an axial direction. The occurrence of cracks and defects at the interface was observed. FIG. 4a is a photograph of the overall cross-section of the sintered duplex nuclear fuel pellet fabricated in Comparative Example 1. FIG. 4b is an optical microscope image showing the vicinity of the interface of the sintered duplex nuclear fuel pellet. As shown in FIGS. 4a and 4b, many cracks were observed in the inner and outer portions of the sintered duplex nuclear fuel pellet. As explained earlier, the occurrence of cracks in the nuclear fuel sintered pellet fabricated in Comparative Example 1 is due to non-uniform densification resulting from a large difference in the sintering rate of the outer and inner portions. Specifically, as shown in FIG. 2, the inner portion of the sintered duplex nuclear fuel pellet composed of the mixed powder UO2-10 wt % Gd2O3 has a maximum densification rate at 1,510° C., whereas the outer portion of the sintered duplex nuclear fuel pellet composed of the mixed powder UO2-2 wt % Er2O3 has a maximum densification rate at 1,220° C. Accordingly, there is a large difference in the contraction at the interface with increasing sintering temperatures. This difference generates a stress at the interface, resulting in the occurrence and growth of cracks. As shown in FIG. 4b, the cracks begin to grow at the interface between the inner and outer portions, and propagate through the inner portion in a direction perpendicular to the interface and through the outer portion in a direction virtually parallel to the interface. This crack propagation behavior suggests that a tensile stress is generated parallel to a lengthwise direction in the inner portion, and a compressive stress is generated parallel to a lengthwise direction in the outer portion. Accordingly, it appears that cracks occur due to a stress generated while the inner portion having a relatively low densification rate is contracted after the outer portion having a relatively high sintering rate is sintered. In contrast, according to Example 1 of the present invention, the sintered duplex nuclear fuel pellet can be fabricated without the occurrence of cracks at the interface between the inner and outer portions by adding manganese oxide (MnO) as a sintering additive to the cylindrical inner portion in order to decrease the difference in the densification rate of the two portions upon sintering. In addition to the manganese oxide (MnO), the sintering additive used herein may be pure manganese, manganese sulfide, manganese fluoride, manganese chloride or the combination thereof. Regardless of the kind of these manganese compounds, sintering promotion effects similar to those in Example 1 are attainable. It is preferable to use manganese compounds having a relatively high melting point. Although the present invention has been described herein with reference to the foregoing examples and the accompanying drawings, it is intended that the scope of the present invention is defined by the claims that follow. Accordingly, those skilled in the art will appreciate that various substitutions, modifications and changes are possible, without departing from the technical spirit of the present invention as disclosed in the accompanying claims. It is to be understood that such substitutions, modifications and changes are within the scope of the present invention. That is, although the sintered duplex nuclear fuel pellet containing UO2 has been described in the foregoing examples and the accompanying drawings, the method of the present invention can be applied to the fabrication of sintered duplex nuclear fuel pellets containing different materials. For instance, although the method of the present invention is applied to the fabrication of another sintered duplex nuclear fuel pellet wherein a portion of the UO2 is replaced with plutonium dioxide (PuO2), which is a nuclear material having the same lattice structure as UO2, the same crack preventing effect can be expected. It should, of course, be understood that this modified embodiment are within the scope and spirit of the invention as disclosed in the accompanying claims. As apparent from the above description, according to the present invention, the sintered duplex nuclear fuel pellet can be fabricated which has few or no cracks at the interface between the cylindrical inner and annular outer portions by adding a small amount of pure manganese or the manganese compound as a sintering additive to the mixed powder UO2—Gd2O3 constituting the inner portion. Since the sintered duplex nuclear fuel pellet having no defects has a high mechanical strength and shows superior heat transfer efficiency, it is expected to greatly improve the operational performance in a nuclear reactor. |
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050376020 | summary | The present invention relates to a facility and method for producing radioisotopes having application to Positron Emission Tomography ("PET"). More particularly, the present invention relates to a system utilizing a relatively small, light-weight Radio Frequency Quadrupole ("RFQ") accelerator for accelerating a beam of .sup.3 He.sup.++ ions to an energy level sufficient to produce desired radionuclides when a selected target material is bombarded with the accelerated beam. BACKGROUND OF THE INVENTION PET is a nuclear medicine procedure for imaging and measuring physiologic processes within the body. It depends upon the distribution into the body of a systematically administered radiopharmaceutical labeled with a radioactive isotope ("radioisotope") that decays through the emission of positrons. This is very distinct from other nuclear imaging techniques such as Computed Tomography ("CT") which measures the distribution of electron density, or Magnetic Resonance Imaging ("MRI") which measures the distribution of protons in the body. There are literally hundreds of possible radiopharmaceuticals that find application to neurology, oncology, and cardiology. PET is typically directed to the study of metabolism processes, blood flow, blood pooling, and receptor sites in the brain. In accordance with PET practice, a radiopharmaceutical (sometimes termed the "labeled compound") is injected into or inhaled by a patient after he or she has been positioned properly relative to an adjacent scanner device. It is the function of the scanner device to detect the gamma-rays that are produced when positrons emitted from the radioisotope annihilate with surrounding electrons. For example, a brain metabolism study might involve the injection of a fluorodeoxy-glucose radiopharmaceutical containing .sup.18 F into the blood stream so that it is taken up in the brain at sites of metabolic activity. When an .sup.18 F nucleus decays it emits a positron which, within a distance of a few millimeters, annihilates with an electron producing two oppositely directed 0.511 MeV gamma-rays. Crystal gamma-ray detectors in the scanner device surrounding the patient's head detect the arrival of the gamma-rays and identify the paths on which they traveled, defining the lines along which the annihilation events occurred. Time-of-flight techniques may also be used to locate the position of the events along the lines. Appropriate electronic circuits and a computer system(s) acquire data during the scan and map the distribution of the annihilation events, which coincide with the presence of the radioisotope. Quantitative evaluation of the function under study, as well as an image for display, are produced as a final product of the PET scan. Radioisotopes are presently generated by accelerating protons to an energy of 12 MeV (or deuterons to an energy of 6 MeV) with a cyclotron. This proton/deuteron beam is extracted from the cyclotron and steered to a target material. Automatic chemical processors convert the target material into basic chemical building blocks, called "precursors", needed to make the radiopharmaceuticals of interest. Some state-of-the-art systems produce the final radiopharmaceutical with the aid of a programmed robot to avoid radiation exposure to a radiochemist. The PET scanner, which resembles a CT scanner in physical appearance, along with the cyclotron, targets, and chemical processors form the basic PET system. Unfortunately, the half-life associated with many radioisotopes of interest to PET applications is very short (on the order of minutes), hence it is not possible to manufacture the radiopharmaceuticals at a manufacturing site and transport them to a patient location. Rather, the patient must travel to the site of the PET system where the needed radioisotopes can be produced and used immediately. Because of the sheer size, mass and expense of building and operating just the cyclotron (which is only one element of a PET system), there are relatively few PET facilities available throughout the world. (At present, it is estimated that there are only about 20 PET facilities in the United States, and about 60-70 worldwide.) Only the largest hospitals are able to afford, support and staff such systems. Thus, the benefits of PET remain available to relatively few. What is needed therefore is a PET system that is more affordable and accessible to a larger number of patients and doctors. There are numerous disadvantages of existing low energy cyclotron-based PET systems. For example, some of the radionuclides are produced using a proton beam, while others are produced using a deuteron beam, therefore some beam switching apparatus is required. While such beam switching apparatus is well known in the art, it adds to the complexity and expense of the system. Further, large amounts of power are required for such systems to operate (e.g., the proton/deuteron cyclotron typically requires 100 kW of power to operate). Also, such systems require enriched target materials if the desired radionuclides are to be efficiently produced by the proton/deuteron beam. Such enriched target materials are not readily available, and are costly to produce. Still further, due to the inherent elliptical cross sectional shape of the proton/deuteron beam, the efficient utilization of the beam in a circular target chamber is made more difficult. Moreover, due to the secondary neutrons that are naturally produced from the proton/deuteron irradiation process, thick shields must be built around the target area to confine such neutron radiation. It is not uncommon, for example, for the target chamber of such systems to be surrounded by concrete walls that are a minimum of four feet thick. This shielding, coupled with the mass and weight associated with the other elements of the system, particularly the cyclotron, results in a system that weighs on the order of 300 tons. Such heavy systems can only be installed on a ground or basement floor, thereby severely restricting those facilities where a cyclotron-based PET system could be installed. All of the above factors combine to make the proton/deuteron cyclotron-based PET systems very expensive to build, operate and maintain. As has been indicated, such expense disadvantageously limits the number of PET systems that are built and operated, thereby making the cyclotron-based PET systems generally inaccessible and/or unavailable to many patients, hospitals and doctors. What is needed, therefore, is a radioisotope production system which can produce sufficient quantities of all of the radioisotopes of interest (.sup.18 F, .sup.11 C, .sup.15 O, .sup.13 N) and minimize some or all of the disadvantages discussed above for existing systems. The present invention advantageously addresses this need. SUMMARY OF THE INVENTION The present invention is directed to a relatively inexpensive PET system that is easy to operate and maintain, and that produces all four of the radionuclides of interest to PET applications. Significantly, the system described herein does not require a cyclotron to generate a proton/deuteron beam. Rather, the PET system of the present invention makes use of a readily available ion source to produce a .sup.3 He.sup.++ beam that is accelerated to around 8 MeV using a Radio Frequency Quadrupole ("RFQ") accelerator. This accelerated .sup.3 He.sup.++ beam is then directed to a conventional, non-enriched target material(s) whereat the four primary radionuclides of interest to PET systems, .sup.18 F, .sup.13 N, .sup.15 O, and .sup.11 C, are efficiently produced. Advantageously, the RFQ accelerator is a small, light-weight device and requires significantly less operating power than does the cyclotron. The RFQ advantageously accelerates ions to a prescribed velocity. The RFQ is thus ideal for accelerating multiply charged ions with masses greater than a single proton mass. This characteristic of the RFQ, in combination with the benefits of using .sup.3 He.sup.++ , rather than protons or deuterons as described below, renders use of a .sup.3 He RFQ as an advantageous and novel technique for producing radioisotopes for PET. Further, the neutron-poor nature of the reaction resulting from a .sup.3 He.sup.++ bombardment of the target material significantly reduces the amount of shielding that is required around the target chamber. Moreover, the generally circular cross section of the .sup.3 He.sup.++ beam allows it to interact with the conventional circular cross-section target material in a more efficient manner than is possible with the elliptical cross-sectional shaped proton/deuteron beam of the cyclotron-based system of the prior art. The reduced shielding requirements, coupled with the small RFQ accelerator and the relatively low power requirements thereof, as well as the efficient use of the target material, makes possible a PET system that not only efficiently generates the needed radionuclides for PET applications, but that also is small, light-weight, affordable, and possibly transportable. Hence, the system can either be readily installed in or possibly transported to the hospitals and other medical facilities where it is needed, thereby making the benefits of PET available to a much larger segment of the world's population. The present invention may thus be summarized as a system for producing radionuclides for use with PET is provided, the system including: a source of ions for producing a .sup.3 He.sup.++ beam at a low energy; a radio frequency quadrupole (RFQ) accelerator for accelerating the low energy .sup.3 He.sup.++ beam to a high energy, and a target system. The target system includes at least one target compound selected to produce at least one desired radionuclide when it is irradiated by the accelerated .sup.3 He.sup.++. beam. This desired radionuclide(s) is then combined, in conventional manner, to produce appropriate precursors which can produce any one of the hundreds of possible radiopharmaceuticals that are used in PET or related applications. Further, the present invention may be characterized as a radioisotope production facility for producing radioisotopes for use with PET. Such a facility includes: RFQ accelerator means for producing a high energy beam of .sup.3 He.sup.++ ions; and means for irradiating a selected target material with the high energy .sup.3 He.sup.++ beam; the target material being selected to produce at least one desired radioisotope when irradiated by the high energy .sup.3 He.sup.++ beam. Still further, the present invention encompasses a method for producing a radiopharmaceutical suitable for use with a PET system. This method comprises the steps of: (a) accelerating a beam of .sup.3 He.sup.++ ions using a RFQ accelerator to a high energy level, e.g., at least 8 MeV; (b) irradiating a target compound with the accelerated .sup.3 He.sup.++ beam to produce at least one desired radionuclide; (c) processing the radionuclide obtained in step (b) to produce a desired precursor containing the radionuclide; and (d) preparing a suitable radiopharmaceutical from the precursor. It is a feature of the present invention to provide a PET system that is small and light weight, thereby allowing the system to be transportable. Another feature of the present invention is to provide such a system that operates on roughly 1/5 of the operating power required by the cyclotron-based PET systems of the prior art. A further feature of the invention is to provide a PET system that occupies only about 1/3 of the floor space that is occupied by the cyclotron-based PET systems of the prior art, and that weighs only about 1/10 of what such prior art cyclotron-based systems typically weigh. Yet another feature of the invention is that the single beam used therein, can be readily and inexpensively generated from a commercial source of ions. A further feature of the invention provides a system as above-described that is very simple to operate, typically requiring the operation of only a few push-buttons, thereby requiring minimal training for its operation. This feature is important because a major part of the cost of the current cyclotron-based PET systems is the cost of the staff. When technicians instead of accelerator experts and radiochemists are used to operate the system, a substantial saving in operating costs results. Another feature of the invention contributing to its simplicity is the lack of a beam extraction system. That is, no extraction system is required to extract the .sup.3 He.sup.++ beam from the RFQ accelerator as is required to extract a proton/deuteron beam from a cyclotron. Still another feature of the invention allows the presently available and medically-proven and accepted target systems, including the programmable robotic features thereof, e.g., those used in existing cyclotron-based PET systems, to be used therewith. Significantly, however, due to the neutron-poor nature of the .sup.3 He.sup.++ beam and resulting reactions, no shielding around the accelerator and little shielding around the target chambers is required relative to existing cyclotron-based PET systems. |
claims | 1. A transmitter device comprising:a neutron detector structured to detect neutron flux;a capacitor electrically connected in parallel with said neutron detector;a gas discharge tube comprising an input end and an output end, said input end being electrically connected with said capacitor; andan antenna electrically connected to said output end, said antenna being structured to emit a signal corresponding to the neutron flux. 2. The transmitter device of claim 1 further comprising an oscillator circuit electrically connected with said output end and said antenna; and wherein said oscillator circuit is structured to pulse said antenna. 3. The transmitter device of claim 2 wherein said oscillator circuit comprises a second capacitor and an inductor electrically connected to said second capacitor; and wherein said second capacitor and said inductor are each electrically connected with said antenna. 4. The transmitter device of claim 3 wherein said oscillator circuit further comprises a resistance temperature detector electrically connected in series with said inductor; and wherein said resistance temperature detector is structured to alter the signal emitted by said antenna. 5. The transmitter device of claim 3 wherein said oscillator circuit further comprises a second inductor electrically connected in series with said inductor; and wherein said second inductor is structured to alter the signal emitted by said antenna. 6. The transmitter device of claim 5 wherein said second inductor is a variable inductor. 7. The transmitter device of claim 1 further comprising a number of Marx bank stages electrically connected between said neutron detector and said gas discharge tube. 8. The transmitter device of claim 1 wherein said transmitter device is devoid of a semiconductor. 9. The transmitter device of claim 1 wherein said transmitter device comprises only one single powering mechanism; and wherein said one single powering mechanism is said neutron detector. 10. A nuclear reactor system comprising:a fuel assembly having an instrument thimble; anda transmitter device comprising:a neutron detector disposed within said instrument thimble, said neutron detector being structured to detect neutron flux,a capacitor electrically connected in parallel with said neutron detector,a gas discharge tube comprising an input end and an output end, said input end being electrically connected with said capacitor, andan antenna electrically connected to said output end, said antenna being structured to emit a signal corresponding to the neutron flux. 11. The nuclear reactor system of claim 10 wherein said transmitter device further comprises an oscillator circuit electrically connected with said output end and said antenna; and wherein said oscillator circuit is structured to pulse said antenna. 12. The nuclear reactor system of claim 11 wherein said oscillator circuit comprises a second capacitor and an inductor electrically connected to said second capacitor; and wherein said second capacitor and said inductor are each electrically connected with said antenna. 13. The nuclear reactor system of claim 12 wherein said oscillator circuit further comprises a resistance temperature detector electrically connected in series with said inductor; and wherein said resistance temperature detector is structured to alter the signal emitted by said antenna. 14. The nuclear reactor system of claim 12 wherein said oscillator circuit further comprises a second inductor electrically connected in series with said inductor; and wherein said second inductor is structured to alter the signal emitted by said antenna. 15. The nuclear reactor system of claim 14 wherein said second inductor is a variable inductor. 16. The nuclear reactor system of claim 10 wherein said transmitter device is devoid of a semiconductor. |
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