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044118589	claims	1. In a method of monitoring power developed in a nuclear reactor by generating power measurement signals in response to fission products produced by groups of fuel rods and processing said signals together with data relating to a plurality of parameters associated with the nuclear reactor to produce a precision power shape readout that is intermittantly interrupted, the improvement residing in the selection of sensors which internally generate heat in response to gamma radiation resulting from said fission products whereby the measurement signals are directly related to linear power generation rate of the fuel rods; separately processing said signals without said data to produce a second but continuous power shape readout that is available during the interruptions in the first mentioned precision power shape readout; and calibrating the separately processed signals in accordance with the precision power shape readout to correct the second continuous power shape readout. 2. The improvement as defined in claim 1 wherein said signals are initially calibrated to reflect the rate of gamma heating of the sensors and are processed to reflect power generated by the fuel rods adjacent to each of the sensors. 3. The improvement as defined in claim 2 including the step of modifying the readouts to compensate for delayed signal response of the measurement signals to changes in power. 4. The improvement as defined in claim 1 including the step of modifying the readouts to compensate for delayed signal response of the measurement signals to changes in power. 5. In a method of monitoring power developed in a nuclear reactor by generating power measurement signals in response to fission products produced by groups of fuel rods, the improvement residing in the steps of: utilizing sensors which internally generate heat in response to gamma radiation resulting from said fission products whereby the measurement signals have a delayed response to changes in power; processing said signals to produce a power shape readout; and filtering said readout to compensate for the delayed response to changes in power. 6. The improvement as defined in claim 5 wherein said signals are initially calibrated to reflect the rate of gamma heating of the sensors and are processed to reflect power generated by the fuel rods adjacent to each of the sensors.
	abstract	A composition of matter that experiences an increase rate of radioactive emission is presented. The composition comprises a radioactive material and particles having affinity for Hydrogen or its isotopes. When exposed to Hydrogen, the composition's emission rate increases. Methods of production are also presented.
	abstract	An X-ray Talbot capturing apparatus is shown. A radiation source irradiates radiation through a plurality of gratings. A radiation detector captures a moire image. A holder which holds the gratings includes a receiving unit including a receiving surface with a curve and a pressing unit including a pressing surface with a curve. Each grating is held between the receiving surface and the pressing surface and bent in an arc shape with a point of the radiation source as a center. An elastic member is positioned between a first surface of the grating and the pressing surface or a second surface of the grating opposite of the first surface and the receiving surface. An opening is provided in the holder and the elastic member so as not to block radiation irradiated on the grating.
	abstract	An assembly to be inserted into a nuclear reactor, such as a liquid sodium-cooled fast neutron reactor, includes an assembly hollow body of elongate shape along a longitudinal axis X. The wall of the hollow body includes at least one through-opening. The assembly also includes an assembly element inserted into the hollow body. The assembly element includes at least one flexible blade of which the free end is shaped into a clip-fastening hook collaborating in clip-fastening fashion with the through-opening from inside the hollow body, so as to connect the assembly element to the hollow body. The assembly also includes at least one removable structure for locking the flexible blade clip-fastened into the through-opening. The removable locking structure makes it possible to prevent the flexible blade from flexing and thus the removable locking structure makes it possible to lock a connection between the assembly element and the hollow body.
046559914	claims	1. An apparatus for detecting the absence of a spring element from among a fuel bundle for a nuclear reactor comprising a plurality of dimensionally similar grouped fuel rod elements mutually spaced in a predetermined array; said apparatus comprising: a probe adapted to be guided along rows of said array; said probe including at least one pawl capable of moving between a retracted and an extended position; means for biasing said pawl toward said extended position, said biasing means adapted to urge said pawl into contact with said grouped fuel rod elements during probing along said rows; and said probe being dimensional with respect to said spaced fuel rod elements to allow said pawl to assume said extended position only at a location where a spring element is missing; whereby said probe is locked at said location against withdrawal from said array to indicate the absence of said missing spring element. said pawl being supported substantially at said forward end of said arm; and said biasing means urging said pawl outward relative to said arm. an aperture communicating between opposite sides of said arm; a pivot pin transversely positioned in said aperture; said pawl including first and second opposite ends and being pivotably mounted on said pin at said first pawl end; said aperture being dimensioned to completely contain said pawl in said retracted position; and said biasing means including a torsion spring surrounding said pin and urging said second pawl end beyond one side of said arm and toward said forward end of said arm. said pawl being constrained from pivoting beyond substantially 90.degree. in said forward direction; and said pawl including a rounded forward surface portion at said second pawl end adapted to provide sliding contact with said springs during said probing in said forward direction, and a substantially flat rear pawl surface adjacent said rounded surface portion and angled relative to the latter to define a surface discontinuity therebetween at said second pawl end. said torsion spring urging respective second pawl ends beyond opposite sides of said arm; whereby a pair of adjacent rows can be probed simultaneously. said probe comprising: an elongate arm including oppositely positioned forward and holding ends, an aperture at said forward end communicating between opposite sides of said arm; a pivot pin transversely positioned in said aperture; a pair of substantially identical pawls mounted in superposed relationship on said pin; a pair of substantially identical pawls mounted in superposed relationship on said pin; each of said pawls being pivotably disposed on said pin at a first pawl end and being constrained to pivot between a retracted position wherein said pawl is completely contained in said aperture and an extended position substantially 90.degree. with respect to said arm; each of said pawls including a rounded forward surface portion at a second pawl end opposite said first pawl end adapted to provide sliding contact with said springs during probing in said forward direction, each pawl further including a substantially flat rear pawl surface adjacent said rounded surface portion and angled relative to the latter to define a surface discontinuity therebetween at said second pawl end; a torsion spring surrounding said pin and urging each of said second pawl ends toward said extended position and into contact with said helical springs during probing along said rows; said probe being dimensioned with respect to said mutually spaced helical springs to allow said pawls to assume said extended position only at a location where a helical spring is missing; whereby said probe is locked at said location against withdrawal from said array to indicate the absence of said missing helical spring. 2. Apparatus in accordance with claim 1 wherein said probe comprises an elongate arm including oppositely positioned forward and holding ends respectively; 3. Apparatus in accordance with claim 2 wherein said pawl is adapted to pivot with respect to said arm between said retracted and extended positions. 4. Apparatus in accordance with claim 3 and further including: 5. Apparatus in accordance with claim 4 wherein said elements comprise substantially identical helical springs each mounted on a separate vertical shank, said springs being disposed in a linear matrix array accessible to said probe only in a horizontal direction such that said springs are probed sequentially in a forward direction in each row; 6. Apparatus in accordance with claim 4 or 5 wherein said pin carries a pair of substantially identical pawls pivotably mounted thereon at respective first pawl ends; 7. A probe for detecting the absence of a helical spring from a linear matrix array of mutually spaced, substantially identical helical springs each mounted on a separate vertical shank, said springs being obstructed by surrounding structure so as to be accessible to said probe only in sequence in a forward direction in each row of said matrix;
	summary
	abstract	A semiconductor mask correcting device is provided with an image acquiring unit acquiring a mask image, an extraction unit extracting only a main pattern from the mask data, an inspection unit inspecting a defective portion by comparing the extracted main pattern with a main pattern which is obtained from the mask image after a drawing by matching to each other, and a correction unit correcting the defective portion specified by the inspection unit, wherein the extraction unit includes a recognition section recognizing the main pattern and the assist pattern as a figure, a specification section specifying the assist pattern from figures which is recognized on the basis of a predetermined condition, and a main pattern extracting section extracting as the main pattern a figure other than the assist pattern.
	claims	1. A fissile neutron detection system for detecting incident fissile neutrons, said fissile neutron detection system, comprising:an ionizing thermal neutron detector arrangement including an inner peripheral shape that at least substantially surrounds a moderator region for detecting thermal neutrons that exit the moderator region but is at least generally transparent to the incident fissile neutrons, the thermal neutron detector arrangement further including at least one group of thermal neutron detectors with the thermal neutron detectors of each group in a side-by-side relationship and spaced apart in the side-by-side relationship to form a gap between adjacent ones of the thermal neutron detectors;a moderator arrangement disposed within the moderator region for converting the incident fissile neutrons in the moderator region to thermal neutrons which exit the moderator region to then enter the thermal neutron detector arrangement for detection of at least some of the thermal neutrons to produce an electrical current as a detector output with the moderator arrangement having an outer peripheral shape that is at least generally complementary to said inner peripheral shape and the moderator arrangement includes lateral extents such that any given dimension that bisects the lateral extents includes a length that is greater than any thickness of the moderator arrangement transverse to the lateral extents; andat least one side moderator disposed in one of the gaps outside of the moderator region. 2. The fissile neutron detection system of claim 1 further comprising:a group of said side moderators with one of the side moderators positioned in the gap between each adjacent pair of the thermal neutron detectors outside of the moderator region. 3. The fissile neutron detection system of claim 1 wherein the side moderator includes a thickness dimension between the side-by-side adjacent ones of the thermal neutron detectors that is no more than 5 cm. 4. The fissile neutron detection system of claim 3 wherein the thickness dimension of the side moderator is in a range from 1 cm to 5 cm, inclusively. 5. The fissile neutron detection system of claim 1 wherein each thermal neutron detector sealingly contains a readout gas and each thermal neutron detector supports an active sheet material layer in gaseous communication with the readout gas for detecting thermal neutrons that are incident on the active sheet material layer and the active sheet material layers of the group of thermal neutron detectors cooperate to form an arrangement of active sheet material layers that spans at least a majority of said lateral extents of the moderator arrangement such that a majority of the thermal neutrons that exit the moderator arrangement thereafter impinge on the arrangement of active sheet material layers to cause the active sheet material layer arrangement to emit ionizing particles responsive to the thermal neutrons that initiates an avalanche of ions in the readout gas to produce said electrical current. 6. A fissile neutron detection system for detecting incident fissile neutrons, said fissile neutron detection system, comprising:an ionizing thermal neutron detector arrangement including an inner peripheral shape that at least substantially surrounds a moderator region for detecting thermal neutrons that exit the moderator region but is at least generally transparent to the incident fissile neutrons, the thermal neutron detector arrangement further including at least one group of thermal neutron detectors with the thermal neutron detectors of each group in a side-by-side relationship and spaced apart in the side-by-side relationship to form a gap between adjacent ones of the thermal neutron detectors;a moderator arrangement disposed within the moderator region for converting the incident fissile neutrons in the moderator region to thermal neutrons which exit the moderator region to then enter the thermal neutron detector arrangement for detection of at least some of the thermal neutrons to produce an electrical current as a detector output with the moderator arrangement having an outer peripheral shape that is at least generally complementary to said inner peripheral shape and the moderator arrangement includes lateral extents such that any given dimension that bisects the lateral extents includes a length that is greater than any thickness of the moderator arrangement transverse to the lateral extents; andat least one external moderator disposed outside the moderator region proximate to the ionizing thermal neutron detector arrangement. 7. The fissile neutron detection system of claim 6 wherein the external moderator extends at least partially transverse to said lateral extents. 8. The fissile neutron detection system of claim 6 wherein the ionizing thermal neutron detector arrangement is at least partially enclosed by the external moderator. 9. The fissile neutron detection system of claim 6 wherein the external moderator at least partially encapsulates at least a portion of the thermal neutron detector arrangement.
	summary
	claims	1. A system for radiographic tissue density evaluation, said system comprising:a cassette configured to be exposed to an X-ray source and configured to obtain information to perform intensity standardization of a captured radiographic image of a subject, the cassette having a minimally radio-opaque backing with a spatial homogeneous X-ray radiographic signature used to estimate a source-detector geometrical inhomogeneity;a calibration bar with a predetermined radiographic signature on or within the cassette to serve as reference for performing the intensity standardization; anda software program to analyze and display the captured radiographic image. 2. The system of claim 1, wherein the minimally radio-opaque backing is made of acrylic polymer or other radiolucent material, having a uniform thickness. 3. The system of claim 1, wherein the calibration bar further comprises a graduated radio-opacity inset of standard density items to form a reference range for an entire dynamic range of X-ray exposure. 4. The system of claim 1, wherein the display further comprises a color-coded intensity image that forms an intuitive colormap based on the calibration bar. 5. The system of claim 1, wherein the software program is installed on a server or a computer that is located in the location of the X-ray radiography machine. 6. The system of claim 1, wherein the software program is installed on a remote server accessed over the Internet. 7. The system of claim 1, wherein the software program is offered as a software on-demand service in the cloud that is accessed over the Internet.
	abstract	A method and apparatus for forming a chemical hydride is described and which includes a pseudo-plasma-electrolysis reactor which is operable to receive a solution capable of forming a chemical hydride and which further includes a cathode and a movable anode, and wherein the anode is moved into and out of fluidic, ohmic electrical contact with the solution capable of forming a chemical hydride and which further, when energized produces an oxygen plasma which facilitates the formation of a chemical hydride in the solution.
039473201	claims	1. A fuel element for a neutronic reactor comprising a cylindrical fluid-tight container of nickel, a plurality of solid substantially spherical bodies of uranium containing about 93.5% U.sup.235 within the container a coating of nonfissionable, corrosion-resistant material covering each body, the bodies having substantially equal diameters, the diameter of the container being an odd multiple of the diameter of the bodies, the bodies being disposed in layers between the ends of the container each layer having a central body with two orbits about it, the inner orbit having six bodies and the outer orbit having twelve bodies, and sodium within the container, the sodium and the uranium bodies entirely filling the container.
	summary
	claims	1. A performance system for a plurality of members configured as an entity, comprising:a performance monitor system in each of the plurality of members that monitors member specific metrics, the performance monitor system logs performance metric information based on corresponding member specific configuration settings, the performance monitor system comprises a configuration consumer component that determines settings in the member specific configuration settings and logs performance metric information based on the settings; anda first computer that includes a gathering and aggregation system that gathers performance metric data from the plurality of members and aggregates the performance metric data into a unified result set. 2. The system of claim 1, wherein each of the plurality of members have a gathering and aggregation system such that the unified result set can be gathered and aggregated from any of the plurality of members. 3. The system of claim 1, the performance monitor system for each of the plurality of members employing a time aggregation component adapted to aggregate member specific metrics over time. 4. The system of claim 3, the time aggregation component being further operable to aggregate member specific performance metrics data into data of larger time periods and larger resolutions. 5. The system of claim 4, the time aggregation component aggregates member specific performance metrics data into data of larger time periods and larger resolutions by taking one of an average, a minimum, a maximum, a last and a weighted average of performance metrics data of a first time period and first resolution to evaluate performance metric data of a consecutive time period and consecutive resolution. 6. The system of claim 1, the performance gathering and aggregation system further comprising a performance entity aggregation component adapted to gather and aggregate performance metric data values of a particular time period and resolution from the plurality of members based on a time period and time resolution requested by a requestor. 7. The system of claim 6, the performance entity aggregation component being operable to aggregate data performance values having similar data times to form a unified result set over the particular time period and time resolution. 8. The system of claim 7, the performance entity aggregation component aggregating member specific performance metrics into a unified result set by evaluating a single data value for data points of similar data times by taking one of an average, a minimum, a maximum, a last and a weighted average of data of similar data times. 9. The system of claim 1, the plurality of members logs member specific operational metrics to a data store corresponding to that particular member. 10. The system of claim 1, the gathering and aggregation system being further adapted to receive a request from a requester for operational metric information for the entity and return the unified result set back to the requestor. 11. The system of claim 10, the requestor being one of an external process, an internal process, an external consume, a user interface and another entity. 12. The system of claim 1, the first computer being configurable to receive a configuration setting defining the operational metric information to be logged, the first computer replicating the configuration setting to the plurality of members. 13. The system of claim 12, wherein any of the plurality of members are configurable to receive a configuration setting defining the operational metric information to be logged. 14. The system of claim 1, the performance gathering and aggregation system being operable to aggregate valid operational metric data and compensate for invalid operational metric data. 15. The system of claim 1, the gathering and aggregation system being further adapted to provide a unified result set of operational metric data for a single member. 16. A system for monitoring performance metrics of a plurality of members configured as an entity, comprising:a first computer having configurable performance metric settings for determining performance metrics to be monitored; andeach of the plurality of members of the entity having member specific configuration settings wherein selection of performance metrics in the first computer is propagated to the member specific configuration settings of each of the plurality of members, at least one of the plurality of members includes a performance monitor system that logs performance metric information based on corresponding member specific configuration settings, the performance monitor system comprises a configuration consumer component that determines settings in the member specific configuration settings and logs performance metric information based on the settings. 17. The system of claim 16, wherein the first computer is a first member of the plurality of members and changes to the configurable performance metric settings at the first member are dynamically updated at the member specific configuration settings of the plurality of members. 18. The system of claim 16, the configuration consumer component being notified of changes in the member specific configuration settings and being operable to access these changes through a configuration store. 19. The system of claim 16, the configuration consumer component being operable to access a configuration store to create a global list containing performance metrics to be logged to a data store. 20. The system of claim 19, the performance monitor system further comprising a metric consumer component communicatively coupled to the configuration consumer component wherein the metric consumer component accesses the global list and retrieves performance metric data from a metric source based on the performance metrics in the global list and logs the performance metric data to the data store. 21. The system of claim 20, wherein the configuration consumer component defines a time period for the metric consumer component to retrieve performance metric data from the metric source and log the performance metric data to the data store. 22. The system of claim 21, the performance metric data being logged based on a predefined time period to the data store. 23. The system of claim 22, further comprising a member time aggregation component operable to dynamically aggregate the performance metric data being logged based on a predefined time period and time resolution in the data store to data sets of larger time periods and larger time resolutions. 24. A method for gathering and aggregating performance metrics of a plurality of members configured as an entity, comprising the step of:determining a plurality of settings that correspond to a plurality of member specific configuration settings;employing a configuration consumer component that monitors and logs performance metrics at a plurality of members based upon the settings that correspond to a plurality of member specific configuration settings;querying at least one operational metric from the plurality of members;aggregating the at least one performance metric from the plurality of members to form a unified result set; andsetting a configuration at one of the plurality of members defining the performance metric data to be logged at each of the plurality of members and replicating the configuration to each of the plurality of members. 25. The method of claim 24, further comprising the steps of receiving a request from a requestor for a performance metric for the entity prior to the step of querying and returning the unified result step. 26. The method of claim 24, further comprising the step of aggregating the performance metric over time at each of the plurality of members after the step of monitoring and prior to the step of querying. 27. The method of claim 26, the step of aggregating the operational metric over time at each of the plurality of members comprising the step of aggregating the operational metric into data sets of at least one larger time period and larger time resolution. 28. The method of claim 27, the step of aggregating the operational metric from the plurality of members to form a unified result set comprising aggregating data performance values having similar data times to form a unified result set. 29. A method for monitoring performance metrics of a plurality of members configured as an entity, comprising:determining configurable performance metric settings for determining performance metric types to be monitored;employing a configuration consumer component that monitors and logs performance metric data at each of the plurality of members based on any changes in the performance metric settings;propagating the performance metric settings to a plurality of remaining members of the entity to establish performance metric configuration settings at the plurality of members; anddynamically updating the logging of performance metric data at each of the plurality of members based on any changes in the performance metric settings. 30. The method of claim 29, further comprising an act of logging performance metric data at predefined time periods and resolutions at each of the plurality of members based on the performance metric configuration settings at each of the plurality of members.
	claims	1. An adjusting device of an apparatus for generating a beam of charged particles, wherein said beam is for interacting with a target and wherein said adjusting device comprises:interface means for receiving, from a user of the apparatus, a set of desired values of characteristics of the beam of charged particles;means for determining a set of nominal values of adjusting parameters of the apparatus, corresponding to said characteristics of the beam of charged particles, and for passing said set of nominal values to the apparatus;means for continually or periodically measuring said adjusting parameters of the apparatus, and for computing corresponding values of said characteristics of the beam of charged particles; andmeans for determining, on the basis of the differences of said computed values from said desired values of characteristics of the beam of charged particles, whether a correction of said values of adjusting parameters is necessary. 2. An adjusting device according to claim 1, further comprising means for determining changes of said values of adjusting parameters suitable for restoring the desired values of said characteristics of the beam of charged particles. 3. An adjusting device according to claim 2, wherein said means of determining whether a correction of said values of adjusting parameters is necessary are also adapted for determining if operation of the apparatus needs to be interrupted, in the event of instability of said adjusting parameters. 4. An adjusting device according to claim 1, wherein said interface means are also adapted for informing said user when a correction of said values of adjusting parameters is necessary. 5. An adjusting device according to claim 1, wherein said characteristics of the beam of charged particles comprise at least one of the following: spatial resolution on the target, size of the particle beams, electrical current carried by the beam. 6. An adjusting device according to claim 5, wherein said adjusting parameters comprise at least one of the following: optical mode emission current of a particle source, width or standard deviation of a current distribution, energy of the particles, voltages applied to particle focusing lenses, amplitude of a scanning field, correction of stigmatism of the particle beam, a distance between the apparatus and the target. 7. An apparatus for generating a beam of charged particles, comprising an adjusting device according to claim 1. 8. An apparatus to claim 7, wherein said charged particles are chosen between electrons and ions. 9. An apparatus according to claim 7, comprising one of an etching apparatus and a microscope.
	abstract	A conductive transparent probe used in a probe control apparatus for adjusting a distance between the apex of the probe and a sample by vibrating the probe with an vibrator in a direction perpendicular to the axis of the probe is provided. The conductive transparent probe includes: an optical fiber having a taper part at one end; a conductive transparent film formed on the surface of the taper part; a first metal film formed on the surface of the optical fiber other than the taper part; wherein the conductive transparent film and the first metal film are electrically connected, and length and thickness of the first metal film are determined such that the conductive transparent probe vibrates while contacting with the vibrator.
	abstract	A liquid metal cooled nuclear reactor includes a reactor vessel, a containment, an air flow path, and an injection unit. The vessel has a reactor core and a coolant for the reactor core. The containment surrounds an outside of the vessel. The air flow path removes heat by flowing air around the containment. The injection unit injects filler in a gap between the vessel and the containment.
	description	The instant application claims priority from U.S. Provisional Patent Application Ser. Nos. 62/691,178 filed Jun. 28, 2018 and 62/596,311 filed Dec. 8, 2017, the disclosures of which are incorporated herein by reference. The disclosed and claimed concept relates generally to nuclear power equipment and, more particularly, to a detection apparatus usable with a fuel rod and an instrumentation tube of a fuel assembly of a nuclear reactor. In many state-of-the-art nuclear reactor systems, in-core sensors are employed for directly measuring the radioactivity within the core at a number of axial elevations. Thermocouple sensors are also located at various points around the core at an elevation where the coolant exits the core to provide a direct measure of coolant outlet temperature at various radial locations. These sensors are used to directly measure the radial and axial distribution of power inside the reactor core. This power distribution measurement information is used to determine whether the reactor is operating within nuclear power distribution limits. The typical in-core sensor used to perform this function is a self-powered detector that produces an electric current that is proportional to the amount of fission occurring around it. This type of sensor is generally disposed within an instrument thimble within various fuel assemblies around the core, does not require an outside source of electrical power to produce the current, is commonly referred to as a self-powered detector, and is more fully described in U.S. Pat. No. 5,745,538, issued Apr. 28, 1998, and assigned to the Assignee of this invention. Another type of sensor capable of measuring various parameters of the core, and which is typically disposed within the instrument thimbles in various fuel assemblies around the core, is described in U.S. patent application Ser. No. 15/417,504, filed Jan. 27, 2017. This type of sensor employs a transmitter device that includes a self-powered neutron detector structured to detect neutron flux, a capacitor electrically connected in parallel with the neutron detector, a gas discharge tube having an input end and an output end, and an antenna electrically connected to the output end in series with a resonant circuit. The input end of the gas discharge tube is electrically connected to the capacitor. The antenna is structured to emit a signal comprising a series of pulses representative of the intensity of the neutron flux monitored by the self-powered detector. Other core parameters can also be monitored by their effects on altering the values of the inductance and capacitance of the resonant circuit. Still another in-core sensor, one which does not require signal leads to communicate its output out of the reactor, is disclosed in U.S. Pat. No. 4,943,683, which describes an anomaly diagnosis system for a nuclear reactor core having an anomaly detecting unit incorporated into a fuel assembly of the nuclear reactor core, and a transmitter-receiver provided outside the reactor vessel. The transmitter-receiver transmits a signal wirelessly to the anomaly detecting unit and receives an echo signal generated by the anomaly detecting unit wirelessly. When the anomaly detecting unit detects an anomaly in the nuclear reactor core, such as an anomalous temperature rise in the fuel assembly, the mode of the echo signal deviates from a reference signal. Then the transmitter-receiver detects the deviation of the echo signal from the reference signal and gives an anomaly detection signal to a plant protection system. The sensor actually monitors coolant temperature around the fuel assembly in which it is mounted. While each of the foregoing sensors directly monitors conditions within the core of a nuclear reactor, such sensor have not been without limitation. Improvements thus would be desirable. None of the aforementioned sensors directly monitors conditions within a nuclear fuel rod in the core during reactor operation. Before advanced fuel cladding materials can be put into commercial use they have to be rigorously tested to receive regulatory approval. The existing methodology for testing advanced fuel cladding materials requires fuel rods to be tested over several fuel cycles and examined at the end of the irradiation test. This is a lengthy process that takes several years during which time fuel cladding data is not available. In the existing method, critical data is only obtained during the post irradiation examination activities. What is desired is an in-pile sensor that can be placed within a fuel rod, endure the hazardous conditions over several fuel cycles, and does not require penetrations into the cladding of the fuel rod. This invention achieves the foregoing objective by providing a nuclear fuel rod real-time passive integral detection apparatus with a remote inductive or magnetic interrogator (also known as pulse induction). The detection apparatus includes a resonant electrical circuit configured to be supported within an interior of a nuclear fuel rod and structured to generate a generally sinusoidal response pulse in response to an incoming excitation pulse and transmit the response pulse in the form of a magnetic wave that travels through a cladding of the nuclear fuel rod to another location within a reactor in which the nuclear fuel rod is housed, wherein a characteristic of the generated pulse is indicative of a condition of the fuel rod. The detection apparatus also includes a transmitter structured to be positioned outside the cladding, in the reactor, in the vicinity of the fuel rod and configured to generate the excitation pulse and transmit the excitation pulse through the cladding to the resonant electrical circuit, and a receiver structured to be supported within the reactor outside of the cladding, in the vicinity of the nuclear fuel rod, and configured to receive the response pulse and, in response to the response pulse, communicates a signal to an electronic processing apparatus outside of the reactor. Preferably, the resonant circuit is supported within a plenum of the nuclear fuel rod. In one such embodiment the characteristic of the response pulse is indicative of the center-line fuel pellet temperature. In another such embodiment the characteristic of the response pulse is indicative of fuel pellet elongation. In still another such embodiment the characteristic of the response pulse is indicative of fuel rod internal pressure. Furthermore, the characteristic of the response pulse may be configured to be simultaneously indicative of a plurality of conditions of the fuel rod. An additional resonant electrical circuit can also be located in a bottom portion of the fuel rod in order to provide measurements at two different axial locations. Preferably, the resonant circuit comprises a plurality of circuit components whose properties such as capacitance and inductance are selected to create a response pulse having a unique frequency, which can be interpreted to identify the particular nuclear fuel rod from which the generated pulse emanated. In addition, the detection apparatus may include a calibration circuit that is configured to be supported within the interior of the nuclear fuel rod and structured to generate a static calibration signal when interrogated by the excitation pulse from the transmitter, which can be received by the receiver and used to correct the response pulse received by the receiver for any signal change associated with component degradation or temperature drift. Accordingly, an aspect of the disclosed and claimed concept is to provide an improved detection apparatus usable with a fuel rod from among a plurality of fuel rods of a fuel assembly, the fuel rod having a cladding that has an interior region, the fuel rod being situated within a nuclear reactor, the detection apparatus being cooperable with an electronic processing apparatus situated outside of the reactor. The detection apparatus can be generally stated as including a transmitter that is structured to be positioned outside the cladding and inside the nuclear reactor in the vicinity of the fuel rod and structured to generate an excitation pulse and to transmit the excitation pulse through the cladding and into the interior region, an electrical circuit apparatus having a resonant electrical circuit that is structured to be supported within the interior region and to generate a response pulse in response to the excitation pulse and to transmit the response pulse in the form of a magnetic field signal that is structured to travel from the interior region and through the cladding, the resonant electrical circuit comprising a plurality of circuit components, at least one circuit component of the plurality of circuit components having a property which is structured to vary in response to a condition of the fuel rod and which, responsive to a change in the condition, is structured to cause the property and the response pulse to vary with the change in the condition and to be indicative of the condition, and a receiver structured to be supported within the nuclear reactor outside the cladding and in the vicinity of the fuel rod, the receiver being structured to receive the response pulse and to communicate to the electronic processing apparatus an output responsive to the response pulse. Another aspect of the disclosed and claimed concept is to provide an improved method of detecting a condition of a fuel rod from among a plurality of fuel rods of a fuel assembly, the fuel rod having a cladding that has an interior region, the fuel rod being situated within a nuclear reactor. The method can be generally stated as including employing a detection apparatus to detect the condition, the detection apparatus being cooperable with an electronic processing apparatus situated outside of the nuclear reactor, the detection apparatus having a transmitter that is structured to be positioned outside the cladding and inside the nuclear reactor in the vicinity of the fuel rod, an electrical circuit apparatus having a resonant electrical circuit that is structured to be supported within the interior region, the resonant electrical circuit comprising a plurality of circuit components, at least one circuit component of the plurality of circuit components having a property which is structured to vary in response to a condition of the fuel rod and which, responsive to a change in the condition, is structured to cause the property and the response pulse to vary with the change in the condition and to be indicative of the condition, and a receiver structured to be supported within the nuclear reactor outside the cladding and in the vicinity of the fuel rod. The employing can be generally stated as including generating with the transmitter an excitation pulse and transmitting the excitation pulse through the cladding and into the interior region, generating with the resonant electrical circuit a response pulse in response to the excitation pulse and transmitting the response pulse in the form of a magnetic field signal from the interior region and through the cladding, and receiving on the receiver the response pulse and communicating to the electronic processing apparatus an output responsive to the response pulse. Similar numerals refer to similar parts throughout the specification. An improved detection apparatus 4 in accordance with the disclosed and claimed concept is depicted generally in FIG. 1. The detection apparatus 4 is usable with a fuel rod 6 and an instrument thimble 8, such as are included in a fuel assembly 10 (FIG. 2) of a nuclear reactor that is depicted schematically in FIG. 2 at the numeral 12, which signifies a containment of the nuclear reactor 12. The detection apparatus 4 is situated within the containment of the nuclear reactor 12, and the detection apparatus 4 is cooperable with an electronic processing apparatus 16 that is situated external to the containment of the nuclear reactor 12. The detection apparatus 4 is thus intended to be situated within the harsh environment situated within the interior of the containment of the nuclear reactor 12 whereas the electronic processing apparatus 16 is situated in a mild environment external to the containment of the nuclear reactor 12. As can be understood from FIG. 1, the electronic processing apparatus 16 can be seen as including a transceiver 18 and a signal processor 22. The transceiver 18 is connected with a wired connection with an interrogation apparatus 48 that is situated in the instrument thimble 8. The signal processor 22 includes a processor and storage 24, with the storage 24 having stored therein a number of routines 28, and the storage 24 further having stored therein a number of data tables 30. The routines 28 are executable on the processor to cause the detection apparatus 4 to perform various operations, including receiving signals from the transceiver 18 and accessing the data tables 30 in order to retrieve values that correspond with aspect of the signals from the transceiver 18 that are representative of conditions inside the fuel rod 6. As can further be understood from FIG. 1, the fuel rod 6 can be said to include a cladding 32 and to have an interior region 36 situated within the cladding 32 and a number of fuel pellets 38 situated within the interior region 36. As employed herein, the expression “a number of” and variations thereof shall refer broadly to any non-zero quantity, including a quantity of one. The fuel rod has a plenum 42 in generally a vertically upper end of the fuel rod 6. The detection apparatus 4 can be said to include an electrical circuit apparatus 44 that is supported within the plenum 42 of the fuel rod 6 within the interior region 36 thereof. The detection apparatus 4 further includes the interrogation apparatus 48, which can be said to be situated within an interior of the instrument thimble 8. As is schematically depicted in FIG. 1, the electrical circuit apparatus 44 is situated within the interior region 36 and communicates with the interrogation apparatus 48 without any breaches or other openings being formed in the cladding 32, thereby advantageously keeping the cladding 32 intact and advantageously keeping the fuel pellets 38 fully contained within the interior region 36. As can be further understood from FIG. 1, and as will be set forth in greater detail below, the electrical circuit apparatus 44 and the interrogation apparatus 48 communicate wirelessly with one another. Conditions within the interior region 36 of the fuel rod 6 can be said to include a temperature of the fuel pellets 38, an extent of physical elongation of the fuel pellets 38, and the ambient pressure within the interior of the fuel rod 6, by way of example. These three conditions are directly detectable by the electrical circuit apparatus 44 and are communicated through the interrogation apparatus 48 to the electronic processing apparatus 16. As will likewise be set forth in greater detail below, various embodiments are disclosed wherein the temperature and elongation of the fuel pellets 38 are detected in various ways, and wherein the ambient pressure within the interior region 36 of the fuel rod 6 is detected in various ways. It is understood that these properties are not intended to be limiting, and it is also understood that other properties potentially can be detectable without departing from the spirit of the instant disclosure. As can be understood from FIG. 3, the electrical circuit apparatus 44 can be said to include a resonant electrical circuit 50 that operates as a sensor and that includes a plurality of circuit components that include a capacitor 54 and an inductor 56. The circuit components have values or properties, such as the capacitance of the capacitor 54 and the inductance of the inductor 56, by way of example, which are selected to impart to the resonant electrical circuit 50 a unique nominal frequency which, when detected by the interrogation apparatus 48, is usable to identify the particular fuel rod 6 within which the electrical circuit apparatus 44 is situated. In this regard, it is understood that a plurality of instances of the electrical circuit apparatus 44 can be situated in a plurality of corresponding fuel rod 6 of the fuel assembly 10. During operation of the detection apparatus 4, the interrogation apparatus 48 interrogates the electrical circuit apparatus 44 in order to receive a signal from the electrical circuit apparatus 44 that can be interpreted as being indicative of one or more of the properties or conditions within the interior region 36 of the fuel rod 6, such as temperature and/or elongation of the fuel pellets 38, ambient pressure within the interior region 36 of the fuel rod 6, etc., and by way of example. The fuel assembly 10 includes a large number of the fuel rods 6, and a subset of the fuel rods 6 of the fuel assembly 10 are envisioned to each have a corresponding electrical circuit apparatus 44 situated therein. When the interrogation apparatus 48 sends out its interrogation signal, the various electrical circuit apparatuses 44 will responsively output a signal that is transmitted through the cladding 32 or the corresponding fuel rod 6 and is received by the interrogation apparatus 48. The various signals from the various electrical circuit apparatuses 44 each has a unique nominal frequency that is selected by selecting the various properties of the capacitor 54 and the inductor 56, by way of example, of the electrical circuit apparatus 44 in order to provide such a signature frequency. The electric processing apparatus 16 is thus able to use the frequencies of the various detected signals to determine which signal corresponds with which fuel rod 6 of the fuel assembly 10. As can further be understood from FIG. 3, the electrical circuit apparatus 44 additionally includes a resonant electrical circuit 60 that is usable as a calibration circuit. That is, the resonant electrical circuit 50 is usable as a sensor circuit that senses the property or condition within the interior region 36 of the fuel rod 6, and the resonant electrical circuit 60 is usable as a calibration circuit to compensate the signal from the resonant electrical circuit 50 for component degradation, temperature drift, and the like. In this regard, the resonant electrical circuit 60 includes a capacitor 62 and an inductor 66 that are selected to have the same material properties as the capacitor 54 and the inductor 56 of the resonant electrical circuit 50. However, and as will be set forth in greater detail below, the resonant electrical circuit 50 is exposed to the condition that is being measured within the interior region 36, such as the temperature and/or elongation of the fuel pellets 38, and/or the ambient pressure within the interior region 36, by way of example. The resonant electrical circuit 60, being usable as a calibration circuit, is generally not so exposed to the condition being measured. Such calibration is provided by employing a ratiometric analysis such as will be discussed in greater detail elsewhere herein. As can further be understood from FIG. 3, the interrogation apparatus 48 can be said to include a transmitter 68 and a receiver 72. The transmitter 68 is configured to output an excitation pulse 74 which is in the form of a magnetic field signal that is capable of being transmitted through the cladding of the instrument thimble 8 within which the interrogation apparatus 48 is situated and is further capable of being transmitted through the cladding 32 of the fuel rod 6. The excitation pulse 74 is thus receivable by the inductor 56 and the inductor 66 of the resonant electrical circuits 50 and 60, respectively, to induce a resonant current in the resonant electrical circuits 50 and 60 in a known fashion. The induced currents in the resonant electrical circuits 50 and 60 result in the outputting of a response pulse 78 from the resonant electrical circuit 50 and a response pulse 80 from the resonant electrical circuit 60. The responses pulses 78 and 80 are in the form of magnetic field signals, which are not merely radio frequency signals, and which can be transmitted from the electrical circuit apparatus 44 through the cladding 32 and through the cladding of the instrument thimble 8 and thus be received on the receiver 72. The excitation pulse 74 is of a generally sinusoidal configuration. The response pulses 78 and 80 are likewise sinusoidal pulses, but they are decaying sinusoidal signals, and it is noted that FIGS. 5A and 5B depict a pair of traces that are representative of two different response pulses 78. In this regard, the frequency of the response pulse 78 may correlate with one parameters within the fuel rod 6, such as temperature, the peak amplitude of the response pulse 78 may correspond with another parameter within the fuel rod 6, such as elongation of the fuel pellets 38, and a decay rate of the response pulse rate 78 may correlate with yet another parameter within the fuel rod 6, such as ambient pressure within the interior region 36. As such, the response pulse 78 may be correlated with a plurality of parameters or conditions within the interior region 36 of the fuel rod 6 within which the electrical circuit apparatus 44 is situated. The aforementioned ratiometric analysis of the response pulses 78 and 80 typically involves taking a ratio of the response pulse 78 to the response pulse 80 or vice versa, in order to eliminate the effects of component degradation and temperature drift. For instance, the resonant electrical circuits 50 and 60 may degrade over time thus affecting the signal that is output therefrom. Likewise, the signals that are output from the resonant electrical circuits 50 and 60 can vary with temperature of the nuclear reactor 12. In order to compensate for these factors, it is assumed that the resonant electrical circuit 50 and the resonant electrical circuit 60 will degrade at substantially the same rate over time. Furthermore, the resonant electrical circuits 50 and 60 will be exposed to the same gross, i.e., overall, temperature within the interior of the nuclear reactor 12. By taking the ratio of the response pulses 78 and 80, such as the ratio of the frequencies, by way of example, and by using the ratio to look up in the data tables 30 a corresponding value for temperature, elongation, and/or pressure, the individual effects of component degradation and temperature drift in the resonant circuit 50 are eliminated. This is because the ratiometric signal is independent of component degradation and temperature drift since the resonant electrical circuits 50 and 60 are assumed to both experience the same component degradation and temperature drift. As is best shown in FIG. 4, the electrical circuit apparatus 44 further includes a elongation transmission apparatus 84 that is situated within the interior region 36 of the fuel rod 6. The elongation transmission apparatus 84 includes a support 86 that is formed of a ceramic material in the depicted exemplary embodiment and which is abutted against the stack of fuel pellets 38. The support 86 has a receptacle 87 formed therein, and the elongation transmission apparatus 84 further includes an elongated element that is in the form of a ferritic rod 88 and that is received in the receptacle 87. The inductor 56 includes a coil 90 that is situated about and exterior surface of a tube 92 that is formed of a ceramic material. The tube 92 has an interior 94 within which an end of the ferritic rod 88 opposite the support 86 is receivable. As the fuel pellets 38 increase in temperature, they thermally expand, thus causing the fuel pellets 38 to push the support 86 and thus the ferritic rod 88 in a rightward direction in FIG. 4, and thus to be received to a relatively greater extent within the interior 94, which alters the inductance of the inductor 56. Such an alteration of the inductance of the inductor 56 adjusts the frequency of the resonant electrical circuit 50, which is detectable when the excitation pulse 74 excites an electrical resonance in the resonant electrical circuit 50. The response pulse 78 from the resonant electrical circuit 50 thus has a frequency that is indicative of the extent of elongation of the fuel pellets 38. The response pulses 78 and 80 are received by the receiver 72, and the receiver 72 responsively sends a number of signals to the electronic processing apparatus 16. The electronic processing apparatus 16 uses the ratio of the response pulses 78 and 80, or vice versa, to retrieve from the data tables 30 an identity of the fuel rod 6 within which the electrical circuit apparatus 44 is situated, based upon the signature nominal frequency of the response pulses 78 and 80, and additionally retrieves from the data tables 30 a value that corresponds with the extent of elongation of the fuel pellets 38 as exemplified by the response pulse 78. These data can then be sent into a main data monitoring system of the nuclear reactor 12, by way of example, or elsewhere. In this regard, it is noted that the calibration circuit represented by the resonant electrical circuit 60 is not strictly critical for the detection of the properties or conditions such as fuel elongation, center line fuel temperature, and ambient pressure, within the interior of the various fuel rods 6. As such, it is understood that the calibration circuit 60 is optional in nature and is usable in order to simplify the data gathering operation and to overcome limitations associated with component degradation and temperature drift, but the calibration circuit 60 is not considered to be necessary to the operation of the detection apparatus 4. As such, it is understood the various other types of electrical circuit apparatuses in the various other embodiments that are described elsewhere herein may or may not include a calibration circuit without departing from the spirit of the instant disclosure. In this regard, it is noted that the calibration circuit 60 is described only in terms of the electrical circuit apparatus 44, but it is understood that any of the other embodiments of the other electrical circuit apparatuses herein may incorporate such a calibration circuit. As suggested above, the response pulse 78 is a decaying sine wave that has properties such as a peak amplitude, a frequency, and a rate of decay. FIG. 5A depicts a trace 96A of one such response pulse 78, and FIG. 5B depicts another trace 96B of another such response pulse 78. It can be understood from FIGS. 5A and 5B that the trace of FIG. 5A has a greater peak amplitude, a higher frequency (as indicated by the shorter period 98A compared with the period 98B in FIG. 5B), and further has a higher rate of decay than the trace 96B of FIG. 5B. As such, while any one of temperature, pressure, and elongation can be directly measured from the frequency of either of the traces 96A and 96B, it is understood that a plurality of such parameters can be simultaneously derived from each such trace 96A and 96B depending upon the configuration of the routines 28 and the data tables 30, by way of example. It thus can be said that elongation of the fuel pellets 38 can affect the inductance value of the inductor 56 by virtue of the relative movement of the ferritic rod 88 with respect to the coil 90. This affects the frequency of the response pulse 78 that is output by the resonant electrical circuit 50, and which is therefore detectable by the electronic processing apparatus 16 through the use of the routines 28 and the data table 30. FIG. 6 depicts an improved electrical circuit apparatus 144 in accordance with a second embodiment of the disclosed and claimed concept. The electrical circuit apparatus 144 includes a resonant electrical circuit 150 having a capacitor 154 and an inductor 156, and is thus similar in that fashion to the electrical circuit apparatus 44. However, the electrical circuit apparatus 144 includes a temperature transmission apparatus 184 that enables measurement of the center line fuel pellet temperature within the fuel rod 6. Specifically, the temperature transmission apparatus 184 includes a modified fuel pellet 186 that is modified to have a receptacle 187 formed therein. The temperature transmission apparatus 184 further includes a tungsten rod 189 that is an elongated element and that is received in the receptacle 187. While the elongated element 189 is depicted in the exemplary embodiment described herein as being formed of tungsten, it is understood that any of a wide variety of other refractory metals and alloys such as molybdenum and the like can be used in place of tungsten. The temperature transmission apparatus 184 further includes a ferritic rod 188 that is abutted against the tungsten rod 189, it being understood that the tungsten rod 189 is abutted with the modified fuel pellet 186. The inductor 156 includes a coil 190 that is situated directly on the ferritic rod 188. During operation, the heat that is generated by the fuel pellets 38 and the modified fuel pellet 186 is conducted through the tungsten rod 189 and thereafter through the ferritic rod 188, thereby causing the temperature of the ferritic rod 188 to correspond with the temperature of the fuel pellets 38 and the modified fuel pellet 186. The permeability of the ferritic rod 188 changes as a function of temperature, and the change in permeability with temperature is depicted in a graph that is shown generally in FIG. 7. A portion of the graph of FIG. 7 is encircled and demonstrates the temperature that is typically seen by the ferritic rod 188 after the heat from the modified fuel pellet 186 is transferred to the ferritic rod 188 by the tungsten rod 189 and demonstrates, due to the steepness of the curve at the indicated location in FIG. 7, the correlation between temperature of the ferritic rod 188 and permeability thereof. The permeability of the ferritic rod 188 which, as noted, varies as a function of temperature, affects the inductance of the inductor 156 with the result that the frequency of the response pulse 78 that is output by the resonant circuit 150 varies directly with the permeability of the ferritic rod 188 and thus with the temperature of the fuel pellets 38 and the modified fuel pellet 186. As such, the temperature of the fuel pellets 38 and the modified fuel pellet 186 can be measured by detecting the response pulse 78 that is output by the resonant electrical circuit 150 through the use of the routines 28 and the retrieval from the data tables 30 of a temperature that corresponds with the detected frequency of the response pulse 78. An improved electrical circuit apparatus 244 in accordance with a third embodiment of the disclosed and claimed concept is depicted in FIG. 8 and is usable in a fuel rod in a fashion similar to the electrical circuit apparatus 44. The electrical circuit apparatus 244 is receivable in the interior region 36 of the fuel rod 6 and includes a resonant electrical circuit 250 and a temperature transmission apparatus 284 that detect the temperature of a set of modified fuel pellets 286. The modified fuel pellets 286 each have a receptacle 287 formed therein. The temperature transmission apparatus 284 includes an amount of liquid metal 291 that is liquid during operation of the nuclear reactor 12. The temperature transmission apparatus 284 further includes a ferritic rod 288 that is engaged with the liquid metal 291 and is buoyantly floated thereon and is receivable in the interior of a coil 290 of an inductor 256 of the resonant electrical circuit 250. The liquid metal 291 expands and contracts with temperature increases and decreases, respectively, of the modified fuel pellets 286. The position of the ferritic rod 288 with respect to the coil 290 is thus directly dependent upon the centerline temperature of the modified fuel pellets 286. Such position of the ferritic rod 288 with respect to the coil 290 affects the inductance of the inductor 256 and therefore correspondingly affects the frequency of the resonant electrical circuit 250. The response pulse 78 that is generated by the resonant electrical circuit 250 thus is receivable by the receiver 72 and is communicated to the electronic processing apparatus 16, and the routines 28 and the data tables 30 are employed to determine a corresponding temperature of the modified fuel pellets 286 and thus of the corresponding fuel rod 6. FIG. 9 depicts an improved electrical circuit apparatus 344 in accordance with a fourth embodiment of the disclosed and claimed concept. The electrical circuit apparatus 344 is usable inside a fuel rod 6 and includes a resonant electrical circuit 350 and a pressure transmission apparatus 385. The pressure transmission apparatus 385 is configured to enable measurement of the ambient pressure within the interior of the fuel rod 6 and includes a support 386 that abuts the stack of fuel pellets 338. The pressure transmission apparatus 385 further includes a ferritic rod 388 and a vessel in the form of a bellows 393 having a hollow cavity 395 and further having a plurality of corrugations 396 formed therein. The hollow cavity 395 is open and is therefore in fluid communication with the interior region of the fuel rod 6. Moreover, an end of the bellows 393 opposite a ferritic rod 388 is affixed to the support 386. The resonant electrical circuit 350 includes a capacitor 354 and further includes an inductor 356 having a coil 390 that is formed about the exterior of a hollow tube 392 having an interior 394 within which a ferritic rod 388 is receivable. The bellows 393 and the ferritic rod 388 are movably received on a support 386 and are biased by a spring in a direction generally toward the fuel pellets 338. As is understood in the relevant art, as the nuclear reactor 12 is in operation, fission gases are produced that include one or more noble gases. Such fission gases increase the ambient pressure within the interior region of the fuel rod 6. Since the hollow cavity 395 is in fluid communication with the interior region of the fuel rod 6, the increased pressure in the interior region 36 bears upon bellows 393 within the hollow cavity 395 and causes the bellows 393 to expand axially, thereby moving the ferritic rod 388 with respect to the coil 390 and thereby affecting the inductance of the inductor 356. An increase in ambient pressure within the interior region 36 of the fuel rod 6 thus expands the bellows 393, thereby resulting in an incremental further reception of the ferritic rod 388 into the coil 390, which results in a corresponding change in inductance of the inductor 356. The corresponding change in inductance of the inductor 356 affects in a predictable fashion the frequency of the resonant electrical circuit 350 and thus likewise affects the frequency of the response pulse 78 that is output by the resonant electrical circuit 350. As a result, when the response pulse 78 from the resonant electrical circuit 350 is received by the receiver 72 and is communicated to the electronic processing apparatus 16, the routines 28 and the data tables 30 are employed to obtain a corresponding value for the ambient pressure within the interior region 36 of the fuel rod 6. Such value for the ambient pressure can then be communicated to an enterprise data system of the nuclear reactor 12. An improved electrical circuit apparatus 444 in accordance with a fifth embodiment of the disclosed and claimed concept is depicted generally in FIG. 10. The electrical circuit apparatus 444 is situated within an interior region 436 of a fuel rod 6 and includes a resonant electrical circuit 450 that includes a capacitor and an inductor 456. The electrical circuit apparatus 444 further includes a pressure transmission apparatus 485 that includes a vessel in the form of a Bourdon tube 493 which, in the depicted exemplary embodiment, includes a hollow tube that is formed in a helical shape. The hollow tube of the Bourdon tube 493 forms a hollow cavity 495, except that an inlet 497 is formed in an end of the Bourdon tube 493 and thus permits fluid communication with the interior of the Bourdon tube 493. More specifically, the electrical circuit apparatus 444 further includes a support 486 in the form of a seal that extends between the edges of the Bourdon tube 493 adjacent the inlet 497 and extends to an interior surface of the interior region 436 of the fuel rod 6. The support 486 thus divides the interior region 436 into a main portion 481 within which a number of fuel pellets 438 are situated and a sub-region 483 within which the Bourdon tube 493 and the inductor 456 are situated. The Bourdon tube 493 is also supported on the support 486. The support 486 resists fluid communication between the main portion 481 and the sub-region 483, except for the inlet 497 which permits fluid communication between the interior of the Bourdon tube 493 and the main portion 481. The pressure transmission apparatus 485 further includes a ferritic rod 488 that is situated on the Bourdon tube 493 at an end thereof opposite the inlet 497. The inductor 456 includes a coil 490, and movement of the ferritic rod 488 in relation to the coil 490 changes the inductance of the inductor 456 such that the frequency of the response pulse 78 that is generated by the electrical circuit apparatus 444 changes corresponding to the ambient pressure within the main portion 481 of the interior region 436. More specifically, as fission gases accumulate in the main portion 481 of the interior region 436, the ambient pressure within the main portion 481 increases, as does the ambient pressure within the hollow cavity 495 of the Bourdon tube 493. Since the sub-region 483 does not experience the increased ambient pressure that is experienced by the main portion 481, and increase in the ambient pressure within the hollow cavity 495 of the Bourdon tube 493 results in expansion of the Bourdon tube 493 and resultant movement of the ferritic rod 488 in the direction of the arrow 499 with respect to the coil 490. This results in a corresponding change in the frequency of the response pulse 78 that is generated by the electrical circuit apparatus 444. It thus can be seen that changes in ambient pressure within the main portion 481 of the interior region 436 result in a change in inductance of the inductor 456 and a corresponding change in the nominal frequency of the resonant electrical circuit 450 and a resultant change in the frequency of the response pulse 78 that is generated by the electrical circuit apparatus 444. When such response pulse 78 is received by the receiver 72, a corresponding signal is communicated to the electronic processing equipment 16, and the routines 28 and the data tables 30 are used to obtain a corresponding value for the ambient pressure within the interior region 436 for output as desired. An improved electrical circuit apparatus 544 in accordance with a sixth embodiment of the disclosed and claimed concept is depicted generally in FIG. 11. The electrical circuit apparatus 544 is similar to the electrical circuit apparatus 444 in that a Bourdon tube 593 is employed as a vessel having a hollow cavity 595. In the electrical circuit apparatus 544, however, the Bourdon tube 593 includes a plug 597 at an end thereof opposite a ferritic rod 588 such that the hollow cavity 595 of the Bourdon tube is not in fluid communication with the interior region 536 of the fuel rod 6, and an increase in ambient pressure within the interior region 536 causes the Bourdon tube 593 to contract. The Bourdon tube 493 is supported on a support 586 in the vicinity of the plug 597, and a contraction of the Bourdon tube 493 due to increased ambient pressure within the interior region 536 thus moves the ferritic rod 588 in the direction of the arrow 599 with respect to the coil 590. The electrical circuit apparatus 544 includes a resonant electrical circuit 550 having a capacitor and an inductor 556, and movement of the ferritic rod 588 with respect to the coil 590 of the inductor 556 changes the inductance of the inductor 556 and thus changes the nominal frequency of the resonant electrical circuit 550. The electrical circuit apparatus 544 thus includes a pressure transmission apparatus 585 that is similar to the pressure transmission apparatus 485, except that the pressure transmission apparatus 585 includes a Bourdon tube 593 whose hollow cavity 595 is not in fluid communication with the interior region 536 and thus contracts in the presence of an increased ambient pressure within the interior region 536. An improved electrical circuit apparatus 644 in accordance with a seventh embodiment of the disclosed and claimed concept includes a resonant electrical circuit 650 having a capacitor 654 and an inductor. The capacitor 654 includes a pair of plates 652A and 652B that are separated by a dielectric material 653. The electrical circuit apparatus 644 is receivable within the interior region 36 of a fuel rod 6 in order to output a response pulse 78 whose frequency is adjusted responsive to a change in ambient pressure within the interior region 36 of the fuel rod 6. More specifically, the dielectric 653 is hygroscopic in nature and is configured to absorb at least some of the fission gases that are generated during operation of the nuclear reactor 12. Such absorption of the fission gases by the dielectric 653 changes the dielectric constant of the dielectric 653, which adjusts the capacitance of the capacitor 654, with a corresponding effect on the frequency of the response pulse 78 that is generated by the resonant electrical circuit 650. As such, a change in the ambient pressure within the interior region 36 of the fuel rod 6 correspondingly affects the capacitance of the capacitor 654 and thus likewise correspondingly affects the frequency of the response pulse 78 that is generated by the resonant electrical circuit 650. When the response pulse 78 is received by the receiver 72, the receiver 72 responsively provides to the electronic processing apparatus 16 a signal which is used by the routines 28 in conjunction with the data tables 30 to obtain and output a value for the ambient pressure within the interior region 36 of the fuel rod 6 within which the electrical circuit apparatus 644 is situated. An electrical circuit apparatus 744 in accordance with an eighth embodiment of the disclosed and claimed concept is depicted generally in FIG. 13 as being situated within an interior region 736 of a fuel rod 6. The electrical circuit apparatus includes a resonant electrical circuit 750 that includes a capacitor 754 and an inductor 756. The electrical circuit apparatus 744 includes a pressure transmission apparatus 785 that includes a support 786 upon which the capacitor 756 is situated in a stationary fashion and further includes a flexible seal 782. More specifically, the capacitor 754 includes a pair of plates 752A and 752B with a dielectric material 753 interposed therebetween. The plate 752A is situated on the support 786, and the flexible seal extends between the plate 752B and an interior surface of the fuel rod 6 to divide the interior region 736 into a main portion 781 within which a number of fuel pellets 738 are situated and a sub-region 783 within which the inductor 756, the plate 752A, the support 786, and the dielectric 753 are situated. The support 786 is rigid but has a number of openings formed therein such that an increase or decrease in the ambient pressure within the main portion 781 will result in movement of the flexible seal 782 with respect to the support 786. The flexible seal 782 thus resists fluid communication between the main portion 781, which is the location where the fission gases are generated, and the sub-region 783. When the main portion 781 experiences a change in the ambient pressure within the main portion 781, this causes the flexible seal 782 and the plate 752B to move with respect to the plate 752A which, being situated on the support 786, remains stationary. The dielectric material 753 is configured to be at least partially flexible in response to movement of the plate 752B with respect to the plate 752A. However, such movement of the plate 752B with respect to the plate 752A results in a change in the capacitance of the capacitor 754. This results in a corresponding change in the frequency of the response pulse 78 that is generated by the resonant electrical circuit 750 as a result of a change in the ambient pressure within the main portion 781. It thus can be understood that a change in ambient pressure within the main portion 781 of the interior region 736 correspondingly changes the frequency of the response pulse 78 that is received by the receiver 72 and which resultantly communicates a signal to the electronic processing apparatus 16. The electronic processing apparatus 16 then employs its routines 28 and its data tables 30 to determine a pressure value that corresponds with the frequency of the response pulse 78 and which is indicative of the ambient pressure within the main portion 781 of the interior region 736. It thus can be seen that various electrical circuit apparatuses are provided that are able to directly measure parameters such as ambient pressure, centerline fuel pellet temperature, and fuel pellet elongation within the various fuel rods 6 of the fuel assembly 10. As noted, any of the electrical circuit apparatuses can include the calibration circuit that is usable to compensate for component degradation and temperature drift. In addition to the direct measurement of the parameters such as centerline fuel pellet temperature, fuel pellet elongation, and ambient pressure within the interior region of the fuel rods 6, it is reiterated that the response pulse 78 in certain circumstances can be analyzed in terms of its peak amplitude, frequency, and rate of decay in order to indirectly and simultaneously indicate a plurality the same parameters of the fuel rods 6. Other variations will be apparent. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular embodiments disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the appended claims and any and all equivalents thereof.
	abstract	Cylindrical inertial confinement fusion reaction chambers are disclosed according to some embodiments of the invention. These chambers can include neutron moderating/absorbing material, radiation absorbing material, and debris collection material. These chambers can also include various injection ports, nozzles, beam ports, sacrificial layers, absorbers, coolant systems, etc. These chambers can be used with directional and/or omni-directional targets.
	summary
	abstract	A nuclear reactor fuel bundle assembly including: a fuel bundle including an array of fuel rods and water rods mounted in an upper tie plate and housed in walls of a channel, and a pore type debris shield mounted at least partially in the channel, above or below the upper tie plate, the shield extending to or over the walls of the channel, whereby deflecting and/or capturing falling debris from entering the fuel assembly, wherein the shield is design to be durable, yet flexible, and porous.
	abstract	A control rod includes a tie-rod, a handle mounted to an upper end portion of the tie-rod, either a connector plate or a fall velocity limiter mounted to a lower end portion of the tie-rod, sheaths having a U-shaped cross-section, welded intermittently to the tie-rod at a plurality of locations in the axial direction of the tie-rod, and having an upper end welded to the handle and a lower end welded to either the connector plate or the fall velocity limiter, and a neutron absorbing member disposed inside each of the sheaths. An upper end of a weld portion located at uppermost position in an axial direction of the tie-rod among a plurality of weld portions between the tie-rod and the sheath is disposed at a position within a range between 0.8 and 13% of total axial length Ls of the sheath below an upper end of the sheath.
	summary
	abstract	A neutron absorbing coating for use on a substrate, and which provides nuclear criticality control is described and which includes a nickel, chromium, molybdenum, and gadolinium alloy having less than about 5% boron, by weight.
	claims	1. An attenuator system for attenuating a radiation beam, comprising:a first attenuating element placed in a path of a radiation beam for attenuation thereof;a second attenuating element also placed in said path of said radiation beam so as to form an attenuation cascade with said first attenuating element;a first positioner operatively connected to said first attenuating element, which moves said first attenuating element along a first direction; anda first processor operatively connected to said first positioner for controlling motion of said first attenuating element;wherein a two-dimensional attenuation distribution of said first attenuating element varies linearly with respect to at least one coordinate, and wherein one of said first and second attenuating elements has two portions with different slopes and cross-sections. 2. The attenuator system according to claim 1, wherein a two-dimensional attenuation distribution of said second attenuating element varies linearly with respect to at least one coordinate. 3. The attenuator system according to claim 1, wherein said first and second attenuating elements form an attenuating cascade, wherein the attenuating cascade has an attenuation distribution depending on a position of said first attenuating element. 4. The attenuator system according to claim 3, wherein the attenuation distribution of the attenuating cascade is generally uniform over an area equal to a cross-section of the radiation beam for a range of positions of said first attenuating element. 5. The attenuator system according to claim 1, further incorporating a second positioner operatively connected to said second attenuating element, which moves said second attenuating element along a second direction; and a second processor operatively connected to said second positioner for controlling motion of said second attenuating element. 6. The attenuator system according to claim 5, wherein said first direction and said second direction are parallel to each other. 7. The attenuator system according to claim 5, wherein said first direction and said second direction are both perpendicular to said radiation beam. 8. The attenuator system according to claim 1, wherein said first attenuating element has a cross-section coplanar with the radiation beam which is triangular in shape and which has an apex with a positive angle slope, and said second attenuating element has a portion with a cross-section coplanar with the radiation beam which is triangular in shape and which has an apex with a negative angle slope. 9. The attenuator system according to claim 8, wherein magnitudes of the positive and negative angle slopes are equal. 10. The attenuator system according to claim 1, further comprising a radiation sensor that senses attenuated radiation that passes through said first and second attenuating elements, said radiation sensor being in operative communication with said first processor, wherein temporal beam modulation is carried out by sensing a beam intensity drift with said radiation sensor and compensating by moving said first attenuating element by said first positioner. 11. The attenuator system according to claim 1, wherein said first and second attenuating elements have cross-sections that vary along a Cartesian coordinate. 12. The attenuator system according to claim 1, wherein said first and second attenuating elements have cross-sections that vary along a polar coordinate. 13. A radiotherapy system comprising:a radiation beam source which emits a radiation beam;a first attenuating element placed in a path of the radiation beam for attenuation thereof;a second attenuating element also placed in said path of said radiation beam so as to form an attenuating cascade;a first positioner operatively connected to said first attenuating element, which moves said first attenuating element along a first direction; anda first processor operatively connected to said first positioner for controlling motion of said first attenuating element;wherein a two-dimensional attenuation distribution of said first attenuating element varies linearly with respect to at least one coordinate, and wherein one of said first and second attenuating elements has two portions with different slopes and cross-sections. 14. The radiotherapy system according to claim 13, further comprising a radiation sensor that senses attenuated radiation that passes through said first and second attenuating elements, said radiation sensor being in operative communication with said first processor, wherein temporal beam modulation is carried out by sensing a beam intensity with said radiation sensor and moving said first attenuating element with said first positioner. 15. A method for attenuating a radiation beam, comprising:placing a first attenuating element in a path of a radiation beam for attenuation thereof;placing a second attenuating element also in said path of said radiation beam so as to form an attenuating cascade; andmoving said first attenuating element along a first direction;wherein a two-dimensional attenuation distribution of said first attenuating element varies linearly with respect to at least one coordinate, and wherein one of said first and second attenuating elements has two portions with different slopes and cross-sections. 16. The method according to claim 15, comprising forming an attenuating cascade with said first and second attenuating elements form, whereas a second positioner is operatively connected to said second attenuating element, which moves said second attenuating element along a second direction; and a second processor is operatively connected to said second positioner for controlling motion of said second attenuating element; and wherein the attenuating cascade has an attenuation distribution depending on a position of said first attenuating element. 17. The method according to claim 15, further comprising carrying out temporal beam modulation by sensing a beam intensity and moving at least one of said first and second attenuating elements.
	description	This is a Continuation-in-part Application of U.S. application Ser. No. 12/187,455, filed on Aug. 7, 2008, which claims the benefit of priority from Korean Patent Application No. 10-2007-0086066, filed on Aug. 27, 2007, the disclosures of both of which are expressly incorporated by reference herein in their entireties. 1. Field of the Invention The present invention relates generally to joint structures between top nozzles and guide thimbles of nuclear fuel assemblies and, more particularly, to a joint structure between a top nozzle and a guide thimble which is configured such that an inner-extension tube is prevented from undesirably rotating when the top nozzle is separated from the nuclear fuel assembly. 2. Description of the Related Art A nuclear reactor refers to a device that is designed to exert artificial control over the chain reaction of the nuclear fission of fissile materials, thereby achieving a variety of purposes such as the generation of heat, the production of radioisotopes and plutonium, the formation of radiation fields, or the like. Generally, enriched uranium that is obtained by raising the ratio of uranium-235 to a range between 2% and 5% is used in a light water nuclear reactor. The uranium is molded into a cylindrical pellet that weighs 5 g and processed into nuclear fuel that is used in a nuclear reactor. Numerous pellets are embedded into a cladding tube made of Zircaloy which is in a vacuum state. Thereafter, a spring and helium gas are put into the tube, and then a top end closure stopper is welded thereon, thereby making a fuel rod. A plurality of fuel rods constitutes a nuclear fuel assembly and is burned in a nuclear reactor by nuclear reaction. FIG. 1 is a schematic view showing a general nuclear fuel assembly. Referring to FIG. 1, the nuclear fuel assembly includes a skeleton and a plurality of fuel rods 1. The skeleton includes a top nozzle 4, a bottom nozzle 5, a plurality of spacer grids 2, a plurality of guide thimbles 3 and a instrument tube 6. The fuel rods 1 are inserted longitudinally into an organized array by the spacer grids 2 in such a manner as to be supported by means of springs (not shown) and dimples (not shown) which are formed in the spacer grids 2. In order to prevent the formation of scratches on the fuel rods 1 and damage to the springs upon assembling the nuclear fuel assembly, lacquer is applied to the surfaces of the fuel rods 1 before the fuel rods 1 are inserted longitudinally into the skeleton of the nuclear fuel assembly. Subsequently, the top and bottom nozzles are secured to the opposite ends of the nuclear fuel assembly, thereby finishing the procedure of assembly of the nuclear fuel assembly. After the lacquer is removed, the following items of the assembled nuclear fuel assembly are tested: the distance between the fuel rods, distortion, dimensions including the length, etc., thus completing the process of manufacturing the nuclear fuel assembly. As shown in FIG. 2, the top nozzle 4 includes a hold-down plate 42, hold-down springs 43, inner-extension tubes 45, outer guide posts 44, and a flow plate 41. Referring to FIGS. 1 and 2, the inner-extension tubes 45 of the top nozzle 4 are connected to the respective guide thimbles 3 so that the nuclear fuel assembly can be firmly fixed in the reactor and the structural stability of the nuclear fuel can be ensured during the burn-up of the nuclear fuel. The top nozzle 4 and the guide thimbles 3 are joined to each other in such a way as to be removably connected to each other, thereby ensuring a path along which the fuel rods 1 can be drawn out when disassembling the top nozzle 4. Disassembly of the top nozzle 4 from the guide thimbles 3 is carried out in a storage tank. A worker must remotely perform the disassembly work to minimize the harm caused by radiation. Accordingly, the joint structure between the top nozzle 4 and the guide thimbles 3 must be designed such that assembly or disassembly between them can be conducted remotely. FIGS. 2 and 3 illustrate a typical method of connecting the guide thimbles 3 with the top nozzle 4. Referring to the drawings, the method of joining the guide thimbles 3 and the top nozzle 4 will be described. As shown in FIG. 2, an external thread is formed on a lower end 451 of each inner-extension tube 45. As shown in FIG. 3, an internal thread is formed on an inner surface of a threaded portion 31 of each guide thimble 3. The top nozzle 4 and the guide thimbles 3 are joined with each other by thread-coupling. An external thread is formed on a lower end of each outer guide post 44. The outer guide posts 44 are threadedly coupled to the flow plate 41. The threaded lower end of each outer guide post 44 is partially welded to the flow plate 41 to prevent the outer guide post 44 from rotating. Furthermore, in order to prevent each inner-extension tube 45 from becoming loose, a head of the inner-extension tube 45 is partially crimped in a radial direction in such a way as to be put in contact with the outer guide post 44. Moreover, the inner-extension tube 45 can be separated from the outer guide post 44 only when torque of more than a specific strength is applied to the head. However, in the state where the inner-extension tube 45 is joined with the outer guide post 44, when the inner-extension tube 45 of the top nozzle 4 is rotated to be separated from the outer guide post 44, since the distance between an outer surface of the inner-extension tube 45 and an inner surface of the outer guide post 44 is too short, it is difficult to rotate the inner-extension tube 45 along the threads if concentricity is not congruous or if foreign substances have gotten stuck between the outer face and the inner face. That is, due to frictional heat generated by the contact surface, the inner-extension tubes 45 and the outer guide posts 44, which are made of stainless steel, are fused together by a cold welding effect, and hence, loosening does not occur. To solve the above-mentioned problems, there have been disclosed U.S. Pat. No. 4,702,883 entitled “Reconstitutable fuel assembly having removable upper stops on guide thimbles”, and U.S. Pat. No. 4,687,630 entitled “Top nozzle and guide thimble joint structure in a nuclear fuel assembly”. In the prior arts, heads of outer guide posts are removed without any inner-extension tube, and processed to have threads so as to minimize the contact surface when the outer guide posts are removed. Furthermore, each guide thimble is threadedly coupled to a threaded portion of a lower end of the corresponding outer guide post. Thus, two threaded coupling portions are respectively formed on upper and lower ends of each outer guide post. Accordingly, when the head of each outer guide post is rotated to remove the top nozzle, since the outer guide post and the head thereof are threadedly-coupled with each other, the thread-coupling between the outer guide post and the guide thimble may become loosened. Hence, in order to prevent the lower end of the outer guide post from becoming loosened, the outer guide post is equipped with a wedge device; however, this has the problem of the assembling and disassembling processes being complicated. Accordingly, the present invention has been made keeping in mind the above problems occurring in the prior art, and an object of the present invention is to provide a joint structure between a top nozzle and a guide thimble which is configured such that when an inner-extension tube head that has been threadedly coupled to an inner-extension tube body is removed from the inner-extension tube body to disassemble the top nozzle, the inner-extension tube body can be prevented from being undesirably removed from the guide thimble. In order to accomplish the above object, the present invention provides a joint structure between a guide thimble and a top nozzle of a nuclear fuel assembly, the guide thimble being coupled to a spacer grid of the nuclear fuel assembly, the top nozzle including: a flow plate located above the guide thimbles, with a coupling through hole formed through the flow plate; an outer guide post coupled at a lower end thereof to the coupling through hole of the flow plate; an inner-extension tube disposed in the outer guide post in such a way that a lower end of the inner-extension tube passes through the coupling through hole of the flow plate; and an inner-extension tube head coupled both to an upper end of the inner-extension tube and to an upper end of the outer guide post, the inner-extension tube head connecting the inner-extension tube and the outer guide post to each other, wherein rotation-preventing means is provided in at least one of a junction between the inner-extension tube and the coupling through hole of the flow plate and a junction among the inner-extension tube head, the inner-extension tube body and the outer guide post, the rotation-preventing means preventing the inner-extension tube body from rotating when the inner-extension tube head is rotated. The present invention is provided to achieve the above-mentioned object. The present invention is characterized by the constructions of an outer guide post 120, an inner-extension tube 150, a flow plate 160 and a guide thimble 3. (Therefore, a hold-down plate and a hold-down spring which are elements of a top nozzle but are not directly related to the present invention will not be explained herein.) Hereinafter, a joint structure among the above elements will be described in detail. FIG. 4 illustrates the joint structure. Referring to FIG. 5, the shape of the outer guide post 120 is that of a hollow cylinder that has open upper and lower ends. An external thread 102 is formed on a predetermined portion of an outer surface of a lower end of the outer guide post 120 and is used to couple the outer guide post 120 to the flow plate 160. The diameter of a predetermined portion of an upper end of the outer guide post 120 larger than those of other portions of the outer guide post 120. Thus, an annular retaining part 104 is formed by this difference in diameter so that the outer guide post 120 can be joined with a hold-down plate (not shown) by means of the annular retaining part 104. As shown in FIG. 6, the inner-extension tube 150 includes an inner-extension tube body 130 and an inner-extension tube head 140. The shape of the inner-extension tube body 130 is that of a hollow cylinder that has open upper and lower ends in the same manner as that of the outer guide post 120. Furthermore, the inner-extension tube body 130 is longer than the outer guide post 120 and is disposed in the outer guide post 120 in such a way that an upper end of the inner-extension tube body 130 is level with that of the outer guide post 120 while its lower end protrudes outwards from that of the outer guide post 120. An external thread 132 is formed on a predetermined portion of the lower end of the inner-extension tube body 130, in other words, on the portion of the inner-extension tube body 130 that protrudes outwards from the outer guide post 120. The external thread 132 is used to couple the inner-extension tube body 130 to the guide thimble 3 which will be explained in detail later herein. The inner-extension tube head 140 is coupled to the upper end of the inner-extension tube body 130. The shape of the inner-extension tube head 140 is that of a cork stopper of which upper and lower ends are the same in shape but are different in diameter. Further, the inner-extension tube head 140 has a hollow structure which is open on upper and lower ends thereof. An internal thread 142 is formed on a predetermined portion of an inner surface of the lower end of the inner-extension tube head 140 so that it is threaded over a circumferential outer surface of an upper end 134 of the inner-extension tube body 130. Here, an annular retaining part 144 is formed by a difference in diameter between the upper and lower ends of the inner-extension tube head 140. The maximum diameter of the annular retaining part 144 is equal to or larger than the diameter of the upper end of the outer guide post 120 so that the inner-extension tube head 140 can be placed on the upper end of the outer guide post 120 rather than being inserted thereinto. Furthermore, the inner-extension tube head 140 has a thin film structure such that it is crimped into a depression of the outer guide post 120, thus preventing the inner-extension tube head 140 from becoming loose. A rotation-preventing surface 136 is formed on the lower portion of the inner-extension tube body 130. The rotation-preventing surface 136 is disposed above the external thread 132 that is formed on the lower end of the inner-extension tube body 130. The rotation-preventing surface 136 is a planar surface formed by cutting out a portion of an annular flange provided above the external thread 132. The rotation-preventing surface 136 functions to prevent the inner-extension tube body 130 from rotating when the inner-extension tube head 140 is separated from the inner-extension tube body 130. As shown in FIG. 7, a coupling through hole 162 is formed at a predetermined position in the flow plate 160 so that the outer guide post 120 is coupled to the inner-extension tube 150 via the screw coupling the upper peripheral surface on the hole 162. An female screw is formed on a circumferential inner surface of the coupling through hole 162. The diameter of the annular retaining part is the same as the outer diameter of the outer guide post 120. A lower surface of the annular retaining part is put into close contact with an upper surface of the annular flange provided on the inner-extension tube body 130. The diameter of an upper end of the coupling through hole 162 is the same as the outer diameter of the outer guide post 120. An internal thread is formed on a circumferential inner surface of the upper end of the coupling through hole 162 so that the lower end of the outer guide post 120 is threaded into the coupling through hole 162. A rotation-preventing portion 166 is formed on a lower end of the inner surface of the coupling through hole 162. The rotation-preventing portion 166 is disposed at a position corresponding to the rotation-preventing surface 136 of the inner-extension tube body 130 and has a polygonal shape corresponding to the shape of the rotation-preventing surface 136. A threaded portion 31 is provided on an upper end of the guide thimble 3 (refer to FIG. 3). The shape of the threaded portion 31 of the guide thimble 3 is that of a hollow cylinder that has an open upper end. The diameter of the threaded portion 31 is the same as that of the inner-extension tube 150. An internal thread is formed on the inner surface of the threaded portion 31. The lower end of the inner-extension tube body 130 is threadedly coupled to the threaded portion 31 of the guide thimble 3. A process of disassembling the top nozzle from the guide thimble will be explained with reference to FIG. 6. First, the inner-extension tube heads 140 are rotated with respect to the corresponding inner-extension tube bodies 130 and removed therefrom. Thereafter, the top nozzle that includes the outer guide posts 120, the flow plate 160, hold-down springs (not shown) and the hold-down plate (not shown) is separated from the nuclear fuel assembly. When each inner-extension tube head 140 is separated from the corresponding inner-extension tube body 130, the thread-coupling between the inner-extension tube body 130 and the threaded portion 31 of the corresponding guide thimble 3 may become loose. However, by virtue of the rotation-preventing surface 136 formed on the inner-extension tube body 130 and the rotation-preventing portion 166 provided on the flow plate 160, the undesirable separation of the inner-extension tube body 130 from the threaded portion 31 of the guide thimble 3 can be prevented. FIGS. 8 through 10 illustrate first and second embodiments of the rotation-preventing surface of the inner-extension tube body and the rotation-preventing portion of the flow plate. Unlike the inner-extension tube body 130 described above, several planar surfaces may be formed on the annular flange of the inner-extension tube body 130 to provide a plurality of rotation-preventing surfaces 135. In detail, as shown in FIG. 8, the rotation-preventing surfaces 135 are formed on the annular flange at positions spaced apart from each other at angular intervals of 90°. Planar cut portions 133 are formed on opposite sides of each rotation-preventing surface 135. The planar cut portions 133 formed between the rotation-preventing surfaces 135 function to prevent interference between fuel rods and the annular flange of the inner-extension tube body when the fuel rods, which are disposed adjacent to the guide thimble in spacer grids of the nuclear fuel assembly, are inserted longitudinally into or drawn out from the nuclear fuel assembly. The second embodiment of the inner-extension tube body 130 is illustrated in FIG. 9. As shown in FIG. 9, rotation-preventing surfaces 138 are formed on an annular flange of the inner-extension tube body 130 at positions spaced apart from each other at angular intervals of 90°. Recesses 139 are formed in the annular flange on opposite sides of each rotation-preventing surface 138 at positions spaced apart from each other by a predetermined distance. The recesses 139 conduct the same role as that of the planar cut portion 133 described in the first embodiment. The flow plate 160 which is coupled to the inner-extension tube body 130 of the first or second embodiment may be used in such a way that flow plates that correspond to the respective flanges of the inner-extension tube bodies of the first and second embodiments are separately provided. However, as shown in FIG. 10A, the lower end of the coupling through hole of the flow plate 160 preferably has a shape corresponding to the annular flange of the inner-extension tube body 130 that is provided only with the rotation-preventing surfaces 135 without any planar cut portion 133. In this case, as shown in FIG. 10B, when the inner-extension tube body 130 is coupled to the flow plate 160, the rotation-preventing surfaces 135 are put into close contact with the flow plate 160, but portions of the annular flange other than the rotation-preventing surfaces 135 are spaced apart from the flow plate 160 rather than making contact with it. Furthermore, the flow plate 160 that has the shape of FIG. 10A can be coupled not only to the inner-extension tube body of the first embodiment but also to that of the second embodiment. As shown in FIG. 10C, the general shape of the annular flange of the inner-extension tube body of the second embodiment is the same as that of the annular flange of the inner-extension tube body of the first embodiment, except for a difference in shape between the planar cut portions 133 and the recesses 139. Therefore, the flow plate of FIG. 10A can be used for the inner-extension tube body of the second embodiment. FIGS. 11 and 12 illustrate a third embodiment with regard to the rotation-preventing surface of the inner-extension tube body and the rotation-preventing portion of the flow plate. In this embodiment, a plurality of rotation-preventing protrusions 137 are provided on the inner-extension tube body 130 and arranged in the circumferential direction at the same positions as those of the rotation-preventing surfaces 136 of the inner-extension tube body 130 described above. Rotation-preventing recesses 167 that have shapes corresponding to the rotation-preventing protrusions 137 are formed in the flow plate 160 at the same positions as those of the rotation-preventing portions 166 of the flow plate 160 described above. The joint structure between the inner-extension tube body and the inner-extension tube head of the inner-extension tube or between the inner-extension tube body and the outer guide post may be embodied by force-fitting rather than thread-coupling (in a fourth embodiment). Referring to FIG. 13, an outer guide post 220 is shaped like a hollow cylinder that has upper and lower ends that are open. An external thread 202 is formed on a predetermined portion of an outer surface of a lower end of the outer guide post 220. The external thread 202 is used to join the outer guide post 220 with a flow plate 160 which will be explained later herein. The diameter of a predetermined portion of an upper end of the outer guide post 220 is larger than at other portions of the outer guide post 220. Thus, an annular retaining part 204 is formed by this difference in diameter so that the outer guide post 220 can be joined with a hold-down plate (not shown) by means of the annular retaining part 204. A coupling groove 206 is formed in an inner surface of the outer guide post 220. The coupling groove 206 is used to join the outer guide post 220 with an inner-extension tube which will be explained later herein. The inner-extension tube includes an inner-extension tube body 230 and an inner-extension tube head 240. As shown in FIG. 14, the shape of the inner-extension tube body 230 is that of a hollow cylinder that has open upper and lower ends in the same manner as that of the outer guide post 220. Furthermore, the inner-extension tube body 230 is longer than the outer guide post 220 and is disposed in the outer guide post 220 in such a way that an upper end of the inner-extension tube body 230 is level with that of the outer guide post 220 while its lower end protrudes outwards from that of the outer guide post 220. An external thread 232 is formed on a predetermined portion of the lower end of the inner-extension tube body 230, in other words, on the portion of the inner-extension tube body 230 that protrudes outwards from the outer guide post 2220. The external thread 232 is used to join the inner-extension tube body 230 with the guide thimble 3 which will be explained in detail later herein. Furthermore, a circumferential coupling protrusion 234 is provided on the inner-extension tube body 230. The coupling protrusion 234 is disposed on a circumferential outer surface of the inner-extension tube body 230 at a position corresponding to the coupling groove 206 formed in the inner surface of the outer guide post 220 so that when the inner-extension tube body 230 is inserted into the outer guide post 220, they are coupled to each other by the coupling groove 206 and the coupling protrusion 234. A plurality of longitudinal slits 236 of a predetermined length are formed in the upper end of the inner-extension tube body 230. The longitudinal slits 236 make it possible for the upper end of the inner-extension tube body 230 to move elastically so that when the inner-extension tube body 230 is inserted into the outer guide post 220, the coupling protrusion 234 can be easily hooked into the coupling groove 206. Further, a coupling groove 238 is formed in an inner surface of the inner-extension tube body 230. The coupling groove 238 is used to join the inner-extension tube body 230 with an inner-extension tube head 240, and it will be explained later herein. A rotation-preventing surface 239 is formed on the lower portion of the inner-extension tube body 230. The rotation-preventing surface 239 is disposed above the external thread 232 that is formed on the lower end of the inner-extension tube body 230. The rotation-preventing surface 239 is a planar surface formed by cutting out a portion of an annular flange provided above the external thread 232. The rotation-preventing surface 239 functions to prevent the inner-extension tube body 230 from rotating when the inner-extension tube head 240 is separated from the inner-extension tube body 230. As shown in FIG. 15, the inner-extension tube head 240 is coupled to the upper end of the inner-extension tube body 230 and has the shape of a cork stopper whose upper and lower ends are the same in shape but different in area. The inner-extension tube head 240 has a hollow structure which is open on upper and lower ends thereof. A coupling protrusion 242 extends in a circumferential direction and is provided on a circumferential outer surface of a lower portion of the inner-extension tube head 240. The coupling protrusion 242 is disposed at a position that corresponds to the position of the coupling groove 238 of the inner-extension tube body 230 when the inner-extension tube head 240 is joined with the inner-extension tube body 230. A fitting protrusion 244 which extends in a circumferential direction is provided on the circumferential outer surface of the lower portion of the inner-extension tube head 240. The fitting protrusion 244 functions to apply pressure to the inner surface of the inner-extension tube body 230 outwards when the inner-extension tube 250 is coupled to the outer guide post 220, thereby strengthening the coupling between them. A plurality of through holes 246 are formed in the circumferential surface of the lower end of the inner-extension tube head 240. The through holes 246 allow a tool or the like to be inserted thereinto to facilitate removal of the inner-extension tube head when disassembling. A taper 248 is formed in the lower end of the inner-extension tube head 240, thus making it easy to insert the inner-extension tube head 240 into the inner-extension tube body 230 or remove it therefrom. The joining between the top nozzle and the guide thimble using the force-fitting structure seldom causes the inner-extension tube body to be rotated. Therefore, this joint structure can be used regardless of the presence of the rotation-preventing surface As described above, in the present invention, a rotation-preventing surface is formed on an inner-extension tube, and a rotation-preventing portion is formed on a flow plate. Thus, when a top nozzle is separated from a guide thimble, the inner-extension tube is prevented from undesirably rotating, thereby preventing the inner-extension tube from becoming loosened from the guide thimble. Therefore, a separate rotation-preventing member is not required. Furthermore, because the area of the contact portion between elements that rotate can be minimized, the assembly or disassembly of the top nozzle can be facilitated, thus reducing the time required to assemble or disassemble the structure. Although the preferred embodiments of the present invention have been disclosed for illustrative purposes, those skilled in the art will appreciate that various modifications, additions and substitutions are possible, without departing from the scope and spirit of the invention as disclosed in the accompanying claims.
	description	The following co-pending applications are hereby incorporated by reference in their entireties: U.S. application Ser. No. 11/409,109, filed Apr. 21, 2006, now U.S. Publication No. 20070247677; and U.S. application Ser. No. 11/513,742, filed Aug. 31, 2006, now U.S. Publication No. 20080055674. The present disclosure relates to an illuminating apparatus used to illuminate hard-copy documents for digital recording, such as in digital scanners, facsimile machines, and digital copiers. In office equipment, such as digital copiers and facsimile machines, original hard-copy documents are recorded as digital data using what can be generally called a “scanner.” In a typical scanner, a document sheet is illuminated and the light reflected from the document sheet is collected by a SELFOC® or a spherical lens and it is recorded by a photosensitive device such as a CCD (charge coupled device) or CMOS (complementary metal oxide semiconductor) array, to be converted to digital image data. In one embodiment, a narrow strip of the document sheet is illuminated as the sheet is moved through a document handler, or the photosensitive device is moved relative to a platen on which the document sheet is placed. One type of illuminator useful in document scanning includes a light-transmissive element that exploits internal reflections to direct light from one or more point sources, such as light emitting diodes (LEDs) to emerge from an exit surface of the element toward a document. Designing an illuminator for a scanner presents challenges in providing, among other aspects, an even illumination along the narrow strip of the document, as well as providing a suitable illumination profile across the narrow strip. The angular distribution of light produced by the illuminator at the document can vary along the illuminated strip, depending upon the illuminator architecture. Irregularities in the illumination level in the illuminated area can result in defects in the image data, particularly in the case of discrete light sources, such as LEDs. While light guides are able to focus light with high efficiency on the imaging area of the platen, it has been found that glossy document surfaces that are uneven, e.g., crumpled or curved from the platen, may specularly reflect the light toward the photosensitive device unevenly, resulting in bright spots in the image. In a practical application of document scanning, specular flare light is created when a combination of conditions (such as a glossy document not lying perpendicular to the optical axis of the imaging lens) enables a portion of the light from the illumination source to specularly reflect into the imaging sensor and add to its signal output. The specular flare light is an undesired addition to the light signal picked up from the light diffused by the document. This creates an artifact that may or may not be objectionable to customers depending upon the conditions. The artifact is usually noticeable on edges along the fast scan direction (potentially both leading and trailing edges). The glossier the document, and the darker the image content, the more the specular flare artifact is enhanced. For uniform sources, such as fluorescent lamps, the specular flare artifact is usually a continuous line; for discrete sources, such as LED arrays, the specular flare artifact is discontinuous and periodic, often appearing as bright spots in the image. According to one aspect, there is provided an optical element for transmitting light emitted from a linear array of light sources. The optical element includes a light-transmissive member, having a refractive index of about 1.4 to about 1.8, and defining an entry surface for disposal near the light sources, the entry surface defining at least one set of prisms, an exit surface, and a section of a DCPC between the entry surface and the exit surface. According to another aspect, there is provided an apparatus for recording an image on a sheet, comprising a linear array of light sources and an optical element for transmitting light emitted from the linear array of light sources. An optical element extends along the linear array of light sources for directing light from the light sources to the sheet, the optical element defining an entry surface disposed near the light sources. The entry surface defines at least one set of prisms. Collection optics receive light reflected from the sheet, the collection optics defining an acceptance angle. Each prism of the set of prisms defines an angle whereby light exiting the optical element is outside the acceptance angle of the collection optics Aspects of the exemplary embodiment relate to an optical element and to a document scanning apparatus or “scanner” which incorporates the optical element. The exemplary embodiment also relates to a method of scanning physical documents for generating scanned images. The documents to be scanned may comprise sheets of paper or other flexible substrate, on which an image or images to be scanned is disposed. The scanner may form a part of an imaging device, such as a stand-alone scanner, a copier, a facsimile machine, or a multifunction device, in which a scanned image is rendered on paper and/or stored in digital form, for example, for display, processing, or transmission in digital form. With reference to FIG. 1, a document scanner includes a platen 10, which may have distinct parts, on which a document sheet 12 can be placed for recording therefrom. Optionally, associated with platen 10 is a document handler (not shown), which sequentially feeds sheets from a multi-page original document. Normally the document 12 lies flat on the platen. It is shown curved in FIG. 1 to indicate specular reflection conditions that might exist at the edge of a book or if the document is crumpled or otherwise curved. A scan head 16 is positioned to illuminate the document and includes an illuminator 18 and a detector 20. The detector includes a photosensitive device 22 and a lens arrangement 24. Light from a linear array of light sources 50 (shown end-on in the Figure) travels through an optical element 30, which will be described in detail below, and illuminates a thin strip of the document. The detector includes a suitable processing device (not shown) for generating an image comprising signals representative of reflected light recorded by the photosensitive device. The photosensitive device 22, which includes one or more linear arrays of photosensors, records the reflected light. The photosensors may comprise solid-state devices, such as CCD (charge coupled device) or CMOS (complementary metal oxide semiconductor) devices. The lens arrangement 24, such as a SELFOC® lens or other microlens arrangement with a predetermined acceptance angle θ, is interposed between the platen 10 and the photosensitive device 22 for focusing the reflected light on the photosensor array. The scan head 16 can be mounted on a moveable carriage 26, for recording light reflected from images on sheets placed on the main portion of platen 10. In general, the carriage translates in direction A, as shown in FIG. 1. FIG. 2 is a perspective view of one embodiment of an optical element, in isolation, for the document scanner of FIG. 1. The optical element 30 is formed from a single piece of light-transmissive material, such as glass or plastic, having a refractive index of about 1.4 to about 1.8 (a typical plastic used in this context has a refractive index of about 1.52). Element 30 defines an entry surface generally indicated as 34 and an exit surface generally indicated as 38. In this embodiment, the cross-section such as 36 of the element 30 defines at least a section of a DCPC (dielectric compound parabolic concentrator), as such an optical element is described in, for example, U.S. patent application Ser. No. 11/409,109, referenced above. In brief, the DCPC relies on total internal reflection to provide a desirable arrangement of beams emerging from the exit surface 38. Defined in the entry surface 34 is a plurality of sets of prisms. FIG. 3 is a sectional view, through line 3-3 in FIG. 2, giving a detailed view of some sets of prisms. (The prisms can effectively be formed by molding or cutting ridges in the body of element 30, creating the prisms between the ridges.) Element 30 extends in a direction corresponding to the linear array 50 of light sources, and each prism is oriented, as shown, in a direction perpendicular to the direction along the linear array. Each set of prisms 40 is placed to be adjacent a light source 50 providing light into entry surface 34. The distance d between the surface of each LED 50 and the set of prisms 40 should be less than 2 mm, and in some instances the LED 50 can contact the prisms. In practical embodiments, the sets 40 of prisms alternate with gap regions 42, which might be painted or otherwise coated with a coating of about 80% reflectivity, with regard to light traveling within element 30. FIG. 4 is a detailed view of a single prism 44 defined by adjacent ridges in a set of prisms 40. The base p of each prism 44 should be about 0.2 mm to about 1.0 mm in length. Each prism 44 defines an angle δ, defined relative to the horizontal line as shown. The angle δ is chosen such that, in case of specular reflection, the light exiting the light guide is outside the collection of the angle of the lens or optics used in the scanner. This angle is typically 48 degrees or higher if an SLA-09 lens is used. For SLA-12 and SLA-20 lenses the δ values are equal to or higher than 52° and 55°, respectively. In case a spherical lens in place of a SELFOC® lens is used, the prism angle will be chosen such that the light coming out of the light guide is outside the acceptance angle of that lens. Because of limitations of the manufacturing process, a prism with very sharp corners may not be possible, but it will have rounded or flat corners. The radius of the corners formed by each prism 44 and between each prism 44 should be as small as possible. The length of each set 40 of prisms along the element 30 should extend beyond the LED 50 corresponding to the set on both sides, and in the present embodiment could be up to four times the length of the LED 50. In overview, the embodiment provides an illumination system whereby specular flare effects, such as from a glossy original image, or an image not in direct contact with platen 10, can be minimized while the diffuse component of light is not affected. In particular, the DCPC, as disclosed in U.S. patent application Ser. No. 11/409,109, maximizes the amount of light transferred from the LED sources to the document plane. The prisms ensure that the light reaching the document plane has angles greater than the acceptance angle of the imaging lens, thus ensuring that any specularly-reflected light is not imaged by the lens. A SELFOC® lens has a very well defined, and not too large, acceptance angle at each point along the lens, so the prism design is relatively uniform. In a practical application, the angle δ on any portion of the prism must be greater than 48°; otherwise light would be transmitted within the acceptance angle of the SELFOC® lens. This means other surface relief structures in the prior art, such as dimples, etc., that have some portion of the surface slope less than the minimum angle for avoiding transmission of light within the acceptance angle of the SELFOC® lens, will not work as efficiently. The claims, as originally presented and as they may be amended, encompass variations, alternatives, modifications, improvements, equivalents, and substantial equivalents of the embodiments and teachings disclosed herein, including those that are presently unforeseen or unappreciated, and that, for example, may arise from applicants/patentees and others.
047724315	claims	1. A process for the immobilization of nuclear waste in a borosilicate glass, wherein: the following are mixed simulataneously: with vigorous stirring, mixing taking place at between about 20.degree. and 80.degree. C., in proportions corresponding to the desired composition of the glass, the said mixture having an acid pH; and the said mixture is dried, calcined at between 300.degree. and 500.degree. C. and then melted. A. simultaneously mixing glass-forming materials in an aqueous system, the ingredients comprising: B. drying the resultant solidified material mixture; C. calcining the dried mixture of Step B at a temperature of about 300.degree. to 500.degree. C.; D. melting the calcined product of Step C to form a melted glass; and E. solidifying the melted glass to form a borosilicate glass that encapsulates the nuclear waste material. 2. The process as claimed in claim 1, wherein mixing is effected with a stirrer which rotates at more than about 500 rpm. 3. The process as claimed in claim 1 in which the mixing is at about 65.degree. to 70.degree. C. 4. The process as claimed in claim 1, wherein the gel precursor is a sol. 5. The process as claimed in claim 1, wherein the gel precursor is alkaline colloidal silica. 6. The process as claimed in claim 1, wherein the gel precursor is acid colloidal silica. 7. The process as claimed in claim 1, wherein the boron compound is ammonium tetraborate. 8. The process as claimed in claim 1, wherein the boron compound is boric acid. 9. A process for immobilizing nuclear waste in the form of a liquid aqueous solution as a waste material, the process comprising the steps of: 10. A process as defined in claim 9 in which other constituents of the final glass are added in Step A, the adding being simultaneous with the glass forming materials, the other constituents comprising a solution of an aluminum compound that forms A1203 in the final glass. 11. A process as defined in claim 10 in which the other constituents comprises solutions of glass-forming compounds that form Na.sub.2 O, ZnO, C.sub.a O and ZrO.sub.2 in the final glass. 12. A process as defined in claim 9 in which Step A is performed at about 65.degree. to 70.degree. C. 13. A process as defined in claim 12 in which the aqueous system of Step A has a pH of about 2.5 to 3.5. 14. A process as defined in claim 9 in which the drying Step B is about 100.degree. to 105.degree. C. 15. A process as defined in claim 9 in which Step C is conducted at about 350.degree. to 450.degree. C. 16. A process as defined in claim 9 in which the silica gel precursor is alkaline colloidal silical that provides a gel to provide the solidified mixture of Step A. 17. A process as defined in claim 9 in which the aqueous solution of 1 and 2 in Step A are concentrated in which the solutions are at least about 75% of their saturation concentrations.
	claims	1. A radiation source, comprising:a free electron laser configured to emit radiation having a wavelength of at most 30 nm; anda device configured to feed back the emitted radiation to the free electron laser, the device comprising:a first optical component configured to couple out of a beam path the emitted radiation; anda second optical component configured to couple into the free electron laser at least a portion of the coupled-out radiation,wherein the first optical component comprises a diffractive optical component configured to separate an order of diffraction of the emitted radiation and send the separated order of diffraction of the emitted radiation to the second optical component without sending at least one other order of diffraction to the second optical component, andwherein the separated out order of diffraction of the emitted radiation comprises the zeroth diffraction order of the emitted radiation. 2. The radiation source of claim 1, wherein the first optical component comprises a grazing incidence mirror. 3. The radiation source of claim 2, wherein the grazing incidence mirror comprises at least one member selected from the group consisting of silicon carbide (SiC), silicon (Si), copper (Cu), ruthenium (Ru), aluminum (Al), and diamond. 4. The radiation source of claim 2, wherein the second optical component comprises a grazing incidence mirror. 5. The radiation source of claim 1, wherein the second optical component comprises a grazing incidence mirror. 6. The radiation source of claim 5, wherein the grazing incidence mirror comprises at least one member selected from the group consisting of silicon carbide (SiC), silicon (Si), copper (Cu), ruthenium (Ru), aluminum (Al), and diamond. 7. The radiation source of claim 1, wherein the diffractive optical component is drivable. 8. The radiation source of claim 1, wherein the diffractive optical component comprises lines perpendicular to a direction of the radiation incident on the diffractive optical component. 9. The radiation source of claim 1, wherein the diffractive optical component comprises lines parallel to a direction of the radiation incident on the diffractive optical component. 10. The radiation source of claim 1, wherein the diffractive optical component comprises a cooling unit. 11. The radiation source of claim 1, wherein the device comprises two grazing incidence mirrors and two normal incidence mirrors. 12. A method, comprising:providing a radiation source according to claim 1;using the FEL to generate radiation;using the diffractive optical component to separate an order of diffraction of the radiation; andcoupling into the FEL at least a portion of the separated order of diffraction of the radiation without coupling into the FEL at least one other order of diffraction of the radiation. 13. The method of claim 12, further comprising driving the diffractive optical component to control the radiation guided to an image field. 14. An illumination system, comprising:a radiation source according to claim 1,wherein the illumination system is selected from the group consisting of a microlithographic illumination system and a metrology illumination system. 15. A system, comprising:an illumination system comprising a radiation source according to claim 1, the illumination system configured to illuminate an object field; anda projection optical unit configured to image the object field into an image field,wherein the system is a microlithographic projection exposure system. 16. A method of using a microlithographic projection exposure system comprising an illumination system and a projection optical system, the method comprising:using the illumination system to illuminate a reticle in an object field with radiation;using the projection optical system to image the reticle onto a light-sensitive material in an image field,wherein the illumination system comprises a radiation source according to claim 1. 17. The radiation source of claim 1, wherein the diffractive optical component comprises a blazed grating. 18. The radiation source of claim 1, wherein the diffractive optical component comprises a grating having a line density of at least 30 lines per millimeter. 19. A radiation source, comprising:a free electron laser configured to emit radiation having a wavelength of at most 30 nm; anda device configured to feed back the emitted radiation to the free electron laser, the device comprising first, second, third and fourth mirrors,wherein:the first mirror comprises a diffractive optical element;the first mirror is configured to separate an order of diffraction of the emitted radiation and send the separated order of diffraction of the emitted radiation to the second mirror without sending at least one other order of diffraction to the second mirror;the separated out order of diffraction of the emitted radiation comprises the zeroth diffraction order of the emitted radiation;the second mirror is configured to reflect at least a portion of the separated order of diffraction of the emitted radiation to the third mirror;the third mirror is configured to reflect at least a portion of the separated order of diffraction of the emitted radiation to the fourth mirror; andthe fourth mirror is configured to couple into the free electron laser at least a portion of the separated order of diffraction of the emitted radiation. 20. The radiation source of claim 19, wherein:the first mirror is a grazing incidence mirror;the second mirror is a normal incidence mirror;the third mirror is a normal incidence mirror; andthe fourth mirror is a grazing incidence mirror. 21. The radiation source of claim 19, wherein the diffractive optical element is drivable.
063046296	abstract	A scanner comprises a conveyor device, a tunnel housing, a bed assembly housing, an isolating device, and one or two analysis devices. The tunnel housing, together with a top portion of the bed assembly housing, forms a substantially enclosed area for analyzing objects, and the bed assembly housing includes most of the components of the conveyor device. A frameless and coverless tunnel construction method clamps the frameless tunnel around the bed assembly housing, and uses a leaded curtain bracket and slot to increase the enclosed area within the tunnel housing for a given scanner width and height. For a coverless construction, the outside of the tunnel is finished cosmetically and any lead shielding is attached to the inside of the tunnel, instead of the outside. The conveyor device comprises a conveyor belt, rollers and a conveyor tracking device having first and second channels formed in first and second rails. The conveyor belt traverses a forward path and a return path in which the first and second rails are preferably provided. The isolating device comprises a first and a second curtain, which extend outward from first and second brackets. The first bracket and curtain are located at an entrance opening to the tunnel housing, and the second bracket and curtain are located at an exit opening of the tunnel housing. The brackets are adapted for insertion into a slot located in a portion of the housing. The isolating device temporarily seals off the substantially enclosed area formed by the tunnel housing and a top portion of the bed assembly housing. The scanning apparatus is light weight, compact, simpler and less expensive to construct and highly reliable in operation.
	description	The present application claims the benefit of: (1) U.S. Provisional Patent Application Ser. No. 60/971,638, entitled “Non-Intrusive Method To Identify Presence Of Nuclear Materials Using Energetic Prompt Neutrons From Photon Induced Fission And Neutron-Induced Fission” which was filed on Sep. 12, 2007 by William Bertozzi and Robert J. Ledoux, and is hereby incorporated by reference, and (2) U.S. patent application Ser. No. 12/139,050, entitled “Non-Intrusive Method To Identify Presence Of Nuclear Materials Using Energetic Prompt Neutrons From Photon-Induced Fission” which was filed on Jun. 13, 2008 by William Bertozzi and Robert J. Ledoux, and is also hereby incorporated by reference. This invention was made with government support under Contract No. N66001-07-D-0025/Delivery Order No. 0001 awarded by the U.S. Navy. The government has certain rights in the invention. This disclosure relates to systems and methods for detecting the presence of fissionable nuclear materials. The systems and methods make use of the distinctive signals provided by the energy and angular distributions of the prompt neutrons produced in neutron induced fission of nuclei. They may be used to detect the presence of actinide nuclei (in particular those with Z greater than or equal to 89, that of actinium). Some of these nuclei are classified as Special Nuclear Materials (SNM) and may be used in weapons of mass destruction such as nuclear explosives and in dirty bombs. Illicit clandestine shipment of nuclear explosives, materials that can be employed in the fabrication of nuclear explosives, and materials that can be employed in the fabrication of dirty bombs may constitute a major threat to the peace and security of the world. Such materials may be secreted and smuggled in cargo or other shipments in various containers including ordinary luggage, crates, vehicles, cargo containers, etc. by terrorists, potential terrorists, terrorist supporters, or others. Effective and efficient methods and systems are required for the reliable, non-intrusive, detection of such contraband materials in ports and in other cargo and shipping locations in order to reduce the risk of successful illicit shipments, without unduly impeding the worldwide flow of cargo in a manner that is disruptive of normal commerce. Accordingly, it is especially important that the detection methods not produce large numbers of false positive detection events. Passive detection methods, as for example gamma spectroscopy of natural decay, have not proven universally effective since many of the materials of interest are not highly radioactive and are relatively easily shielded. X-ray techniques do not readily distinguish between fissionable nuclear materials and innocuous high-Z materials like lead or tungsten that may be legitimately present in cargo. In addition to passive detection, several approaches to detection have been employed, attempted, or proposed using active techniques employing probing beams. In one such active technique, an external neutron source has been used to detect fissionable nuclear materials by detecting induced fission events by the neutron multiplication effect of the fission events. However, it has been difficult to discriminate between the probing neutrons and the fission induced prompt neutrons, especially when the energy of the probing neutrons is as high as the energy of the more energetic prompt neutrons from fission or when large containers are involved. Alternative techniques have induced fission events in fissionable nuclear materials with pulsed external neutron sources, then detecting delayed emission of neutrons by fission products, using time delay, as a means of distinguishing the detected signal from the probing neutrons. This delayed neutron signal is a much weaker signal, and is subject to signal-to-noise ratio problems. It is therefore an object of this disclosure to provide improved systems and methods for detecting fissionable nuclear material in an article with reduced error and ambiguity. It is a further object of this disclosure to provide improved systems and methods for detecting contraband fissionable nuclear materials by improving discrimination of prompt fission neutrons in the presence of noise-contributing factors. Another object of this disclosure is to provide systems and methods for analyzing the energy or an energy spectrum of prompt fission neutrons to detect the presence of fissionable nuclear materials in an article. A still further object of this disclosure is to provide systems and methods for detecting an angular distribution of prompt fission neutrons to detect the presence of fissionable nuclear materials in an article. Yet another object of this disclosure is to provide systems and methods for using an angular distribution of prompt fission neutrons and an energy distribution of prompt fission neutrons to detect the presence of fissionable nuclear materials in an article. The objects set forth above as well as further and other objects and advantages of the present disclosure are achieved by the embodiments described below. A prompt neutron is a neutron emitted immediately after the fission process; it is characterized by being emitted from a fission fragment generally after the fragment has reached a significant fraction of its final velocity, and thus may be referred to as a fully accelerated fragment. The final velocity is imparted to the fragment by the strong Coulomb repulsion between the fission fragments. There are delayed neutrons that arise following the beta-decay of some of the fragments, but these are not considered herein since they are only a small percentage of the neutrons emitted promptly and thus have a negligible effect on the practice of the methods disclosed herein. One of the advantages of utilizing prompt neutrons from neutron-induced fission as a detection technique is that they are produced with approximately 200 times the yield of delayed neutrons; this allows for higher probabilities of detection, lower false positive rates, and faster scan times. The techniques and methods described in U.S. patent application Ser. No. 12/139,050, to which this application claims priority, and which is incorporated herein, make use of the boost in velocity (and thus energy) of a neutron that arises because the neutron is emitted from a rapidly moving nuclear fragment which has been produced by the (γ, f) process. This boost places the neutron in an energy range that will allow for the unambiguous determination of the presence of fissionable nuclei; this energy range is not possible from other processes that could occur with other non-fissionable nuclei such as direct neutron production by photons (γ, n). Additional features of interest are the nucleus-dependent angular distribution of the fragments in the photo-fission process and the prompt neutron energy distributions at various angles. Thus the signature of photon-induced fission is unique. Also, by controlling the incident photon energy used to cause the fission, (γ, n) processes from other nuclei may be reduced in importance or eliminated as a background. Since the process of photon-induced fission is ubiquitous with the actinides, these methods will identify fissionable nuclear materials within a container, in particular those which have Z equal to or greater than 89, that of actinium. That disclosure describes systems and methods for detecting fissile materials by measuring prompt neutron energies and examining prompt neutron energy spectra. The energy spectra of prompt neutrons that originate from photo-fission are readily distinguishable from the energy spectra of neutrons that originate from other processes that may occur in non-fissile materials such as (γ, n). Neutrons at energies greater than E=Eb−Eth, where Eth is the threshold for the (γ, n) process in relevant other heavy non-fissile elements and Eb is the endpoint energy of the incident bremsstrahlung photon beam (or the energy of an incident monochromatic photon beam), indicate with no ambiguity the presence of fissile material in the actinide region. No other photon-induced process can generate neutrons with these energies. Angular distributions of these neutrons reflect the angular distributions of the fission fragments from which they arise: distributions deviating significantly from isotropy indicate the presence of even-even nuclei while almost isotropic distributions indicate the presence of odd-even or even-odd fissile species. (Hereinafter, in the interests of conciseness, “odd-even” shall denote a nucleus with an odd number of nucleons, whether protons or neutrons, and thus the term hereinafter shall encompass both “odd-even” nuclei and “even-odd” nuclei.) Comparison of the energy distribution of the prompt neutrons at different angles also provides potentially useful information about the species present. If the energy distributions at different angles are nearly identical, the isotopes undergoing fission are odd-even; if the energy distributions differ significantly at different angles, the isotopes undergoing photo-fission are even-even. Another signature of photo-fission is the fact that the relative yield of prompt neutrons at different neutron energies (i.e., the shape of the yield curve) does not depend upon the incident photon energy. This is in contrast to other processes such as (γ, n) where the relative yield of neutrons at different energies is strongly dependent on incident photon energy, particularly at the highest energies possible. For a better understanding of that disclosure, together with other and further objects thereof, reference is made to the accompanying drawings and detailed description. The techniques and methods of this disclosure also make use of the boost in velocity (and thus energy) of neutrons that arises because of the emission of those neutrons from a rapidly moving nuclear fragment after a fission event. In this disclosure, the fission event of interest is triggered by an incident neutron. This energy boost in the emitted neutron contributes to a neutron energy range that will allow for the unambiguous determination of the presence of fissionable nuclei; something not available from other processes such as direct neutron production by photons (γ, n), direct production of neutrons by (n,n′) reactions or (n,2n) reactions, for example. An additional feature is the nucleus-dependent angular distribution of the fragments in the neutron induced fission process and the correlation of that distribution to the prompt neutron energy distributions at various angles. Thus the signature of neutron-induced fission is unique as compared to potentially-competing events. Also, by controlling the neutron energy used to cause the fission, background from the neutron generator used to produce neutrons to induce the fission may be reduced in importance or eliminated, which increases the capabilities of the methods. The probability of neutron induced fission depends strongly on the energy of the neutron and this dependence is different for different nuclear species. Since the process of neutron induced fission is ubiquitous with the actinides, these methods will identify fissionable nuclear materials within a container, which have Z equal to or greater than 89, that of actinium, and will also make possible the identification of the particular nuclear species undergoing fission. This disclosure employs systems and methods for the detection of fissile materials by examining the energy spectra of the prompt neutrons that originate from neutron induced fission in comparison to the energy spectra of neutrons from neutron generators and other neutron induced reactions such as (n,n′) and (n,2n) among others. Neutrons can induce fission at energies as low as thermal energies (˜ 1/40 eV) for materials of interest in the construction of nuclear weapons such as 235U, and 239Pu. The cross sections for (n,f) are very large for thermal neutrons for the critical isotopes 235U and 239Pu but only modest for 237Np. However, for higher incident neutron energies, above 1 MeV, all these nuclei present large and usable cross sections for (n,f). In fact, it is the fast neutron fission cross section which is important to making a weapon from fissionable nuclear material. Using a neutron generator in the MeV range will allow the (n,f) reaction with these nuclei and if the incident neutron energy is less than a few MeV it will not provide a background that can be confused with the neutrons from (n,f) because the (n,f) process produces neutrons with higher energies. Neutrons from (n,f) have energies ranging up to 9 MeV or more. Angular distributions of these prompt neutrons reflect the angular distributions of the fission fragments. This is another signature that can be used to distinguish neutrons from (n,f) and the neutrons from a neutron generator. In particular, the fragments from fission induced by neutrons of several MeV are peaked forward and backwards relative to the initiating neutron direction. At lower energies these fragments may be peaked at 90 degrees, depending on the neutron energy and nucleus. The energy distribution of the prompt neutrons from fission at different angles to the direction of the initiating neutrons also indicates the angular distribution of the fragments. The neutron energy distributions of the prompt neutrons from (n,f) will be different in the directions parallel to and perpendicular to the direction to the initiating neutron, reflecting the motion of the fragments and their preferred directionality. Another signature of neutron induced fission is the fact that the relative yield of prompt neutrons at different neutron energies (i.e., the shape of the fission neutron yield in energy) does not depend strongly upon the energy of the initiating neutron energy. This is in contrast to (n,n′) or (n,2n) processes. This disclosure also employs systems and methods for the detection of fissile materials by examining the energy spectra and angular distributions of prompt neutrons resulting from fission induced by neutrons. For a better understanding of the present disclosure, together with other and further objects thereof, reference is made to the accompanying drawings and detailed description. Fission is a complex process that has been the subject of many theoretical and experimental studies. (See generally Bohr and Mottelson, “Nuclear Structure”, 1998, World Scientific Publishing Co. Pte. Ltd. Singapore, and references therein). However, common empirically established features imply certain general regularities of the process independent of nucleus or initiating particle. When fission is spontaneous, initiated by low energy neutrons or by the absorption of photons near the threshold for the (γ, f) process, the dominant mode of fission is the breaking apart of the nucleus into two fragments of unequal masses. These unequal masses are in the regions of nucleon numbers 95 and 140 for 235U and in similar regions for other fissionable nuclei. The fragments are accelerated by the strong Coulomb repulsion of their charges (Z1, Z2) and gain kinetic energy ranging approximately from 160 to 180 MeV, depending on the nucleus undergoing fission. Most of this Coulomb energy is gained in approximately 10−22 sec as the fragments separate by several nuclear diameters. The final fragment velocities correspond to kinetic energies of approximately 1 MeV/nucleon for the light fragment and approximately 0.5 MeV/nucleon for the heavy fragment. The rapidly moving fragments are generally excited and emit prompt neutrons, mostly after they have gained most of the kinetic energy available from the Coulomb repulsion. FIGS. 1A and 1B display the analysis by J. Terrell (“Neutron Yields from Individual Fission Fragments”, Physical Review, Vol. 127, Number 3, Aug. 1, 1962, pages 880-904, and references therein) for the neutron induced fission of 235U and 239Pu. These figures (which correspond to FIGS. 8 and 9 in Terrell) display the asymmetric fragment mass distributions from the neutron-induced fission and the average number of neutrons emitted from the heavy and light fragments, as a function of the mass of the fragments, for 235U and 239Pu. (The symbols ν, νL and νH in FIGS. 1A and 1B denote the average total number of neutrons emitted, the average neutrons emitted from the light fragment and the average neutrons emitted from the heavy fragment, respectively, as a function of fragment mass.) Similar results have been obtained by Terrell for neutron-induced fission of 233U and the spontaneous fission of 252Cf, showing the generality of the phenomena. Many authors have studied the spontaneous fission of 252Cf, including Harry R. Bowman, Stanley G. Thompson, J. C. D. Milton and J. Swiatecki: “Velocity and Angular Distributions of Prompt Neutrons from Spontaneous Fission of 252Cf”, Phys. Rev., Volume 126, Number 6, Jun. 15, 1962, page 2120-2136 and references therein. These authors were able to demonstrate by direct measurement that: a) “The angular distribution (of the neutrons from the spontaneous fission of 252Cf) is strongly peaked in the direction of the fission fragments. The relative intensities in the direction of the light fragment, in the direction of the heavy fragment and at right angles are about 9, 5 and 1 respectively”: and b) “The broad features of the energy and angular distributions are reproduced by the assumption of isotropic evaporation (in the fragment frame of reference) of the neutrons from fully accelerated fragments.” While not the only important conclusions of the Terrell and Bowman works, those quoted and discussed here sustain the general description of spontaneous fission or fission at low energies that is important to the discussion herein. The work of H. W. Schmitt, J. H. Neiler, and F. J. Walter, “Fragment Energy Correlation Measurements for 252Cf Spontaneous Fission and 235U Thermal-Neutron Fission”, Phys. Rev. Volume 141, Number 3, January 1966, Page 1146-1160, provides additional evidence of the features described above. They find that the average total fragment kinetic energies before neutron emission are 186.5±1.2 MeV for the spontaneous fission of 252Ca and 171.9±1.4 MeV for neutron induced fission of 235U. The fragments have substantially all the kinetic energy available from the mutual Coulomb repulsion of the fragments. Both the energy distribution and the angular distribution of the neutrons from fission fragments created by photon-induced fission are relevant to the detection of such neutrons. The case of 232Th reported in C. P. Sargent, W. Bertozzi, P. T. Demos, J. L. Matthews and W. Turchinetz, “Prompt Neutrons from Thorium Photofission”, Physical Review, Volume 137, Number 1B, Jan. 11, 1965, Pages B89-B101 is illustrative. These authors measured the spectra of neutrons from the photo-fission of 232Th at pairs of angles simultaneously, 157 and 77 degrees relative to the photon beam, and 130 and 50 degrees relative to the photon beam. They used bremsstrahlung photons from electrons with kinetic energies of 6.75 and 7.75 MeV. Several subsidiary facts were important in their analysis: 1.) The (γ, n) threshold energy for 232Th is 6.438 MeV. Therefore, the (γ, n) process cannot contribute neutrons of energy greater than 0.31 MeV and 1.31 MeV, respectively at the two energies of the electron beam, 6.75 MeV and 7.75 MeV. Since these neutron energies are achieved only at the end points of the respective bremsstrahlung spectra, there will not be important contributions to the neutron spectra from the (γ, n) process even at neutron energies considerably lower than 0.31 or 1.31 MeV, respectively; and 2.) The fission fragments in photo-fission, (γ, f), are known to have strongly anisotropic angular distributions from 232Th. The distribution is peaked at 90 degrees to the incident photon beam, and the fragment angular distribution is given by I=a+b sin2(θ), where θ is the angle between the incident photon beam direction and the fission fragment direction. The ratio b/a is considerably larger than 1 at the energies discussed herein and remains larger than one even at incident photon energies higher than 9 MeV. (E. J. Winhold, P. T. Demos and I. Halpern, Physical Review, 87, 1139 (1952): and, A. P. Berg, R. M. Bartholomew, F. Brown, L. Katz and S. B. Kowalski, Canadian Journal of Physics, 37, 1418 (1959)). This fragment directionality provides the correlation between neutron angle and neutron energy that results from the velocity boost if the prompt neutrons are emitted from fragments that have their full kinetic energy. The results of analysis of the neutron energy spectra from 232Th (γ, f) are consistent with the following conclusions of Sargent et al: 1.) The fraction of the prompt neutrons that result from emission from other than the fully accelerated fragments is 0.07±0.09; 2.) The prompt neutron angular distributions and energy distributions are consistent with isotropic neutron evaporation with a thermal-type spectrum in the center of mass frame of reference of the moving fragments, where the fragments are moving with their fully accelerated velocities; and 3.) The energy spectrum of the neutrons in the center of mass frame of reference is characterized by an average energy of 1.14±0.06 MeV. There are no significant components of temperature as high as or higher than this average energy. (That is, the ensuing Maxwellian energy distribution, were it applied to a fragment at rest in the laboratory frame of reference without the kinematic boost from the motion of the photo-fission fragments, would not yield many neutrons at the high energies that result from applying the kinematic boost to neutrons emitted in the fragment frame of reference.) FIG. 7 presents angular distributions of prompt neutrons from the fission fragments produced in the (γ, f) process for incident photon energies near the threshold for the (γ, f) process, for 232Th and 238U. It is taken from S. Nair, D. B. Gayther, B. H. Patrick and E. M. Bowey, Journal of Physics, G: Nuclear Physics, Vol. 3, No. 7, 1977 (pp 1965-1978), who corroborate the relevant 232Th results of Sargent et al. and also extend the results to the photo-fission of 238U. These angular distributions are measured by detectors which detect the fragments from neutron induced fission of 238U. Therefore, they are an average over all the energies of the neutrons emitted from the photo-fission fragments convoluted with the (n, f) cross section. This emphasizes neutrons above approximately 1 MeV, where the (n, f) cross section becomes large (See FIG. 11, which presents the (n, f) cross section for 238U. FIG. 11 is reproduced from National Nuclear Data Center, Brookhaven National Laboratory, ENDF, Evaluated Nuclear (reaction) Data File). FIG. 10, also taken from Nair, presents the angular distributions of the fragments from the photo-fission for 232Th and 238U, for the same incident photon energies as FIG. 7. The peaking of the neutrons from the photo-fission fragments in the direction of the motion of the fission fragments is clearly demonstrated by a visual comparison of FIG. 7 with FIG. 10. (The implications of the shape of the neutron angular distribution are discussed below.) FIG. 2, which is taken from FIG. 4 of the Sargent et al. reference, displays the time-of-flight (energy) spectrum of prompt neutrons from photo-fission of 232Th at 77 degrees with respect to the direction of an incident 7.75 MeV. photon beam. At the top of FIG. 2 is the prompt neutron energy scale. One outstanding feature of the neutron spectrum in FIG. 2 is the presence of neutrons at very high energy compared to an evaporation (thermal) spectrum with an average energy of approximately 1.14 MeV, as reported by Sargent et. al. from their analysis of the energy and angular distributions of the prompt neutrons from photon induced fission of 232Th. For example, the intensity at 6 MeV is considerable. The presence of a large number of neutrons at high energy results in part from the considerable boost in velocity transferred to the neutrons by the moving fragments. For example, if the velocity of the fragment corresponds to a kinetic energy of 1 MeV/nucleon, then a 1 MeV neutron emitted in the fragment center of mass frame of reference in the direction of fragment motion will have twice the velocity in the laboratory frame of reference and a kinetic energy of 4 MeV. This follows because the neutron velocity in the laboratory frame is the sum of the fragment velocity and the neutron velocity in the fragment frame. Since these are the same for the energies and directions considered in this example, the velocity is doubled. The kinetic energy varies as the square of the velocity. Hence the neutrons with 1 MeV in the fragment frame of reference have 4 MeV in the laboratory frame of reference. More generally, if the fragment velocity is V and the co-directional neutron velocity in the fragment frame is v, then the neutron velocity in the laboratory frame is V+v. The kinetic energy of the neutron in the laboratory frame is E=(m/2)(V2+2Vv+v2) or E=Ef(1+2(En/Ef)0.5+En/Ef) where En is the neutron kinetic energy in the fragment frame and Ef is the kinetic energy of one nucleon of the fragment. Thus, in the above example, a neutron emitted in the fragment direction of motion at 2 MeV in the fragment center of mass frame of reference will have a laboratory kinetic energy of 5.8 MeV. Energy conservation in the direct (γ, n) neutron production process does not allow the production of neutrons with an energy above E=Eb−Eth, where Eb is the bremsstrahlung endpoint energy of the incident photon beam and Eth is the (γ, n) threshold energy for producing neutrons from other relevant heavy elements. Therefore, detecting neutrons with energies above this value is definitive evidence of the presence of fission. Since the (γ,n) threshold of 232Th is 6.438 MeV, a neutron energy of 6 MeV will not be possible from (γ, n) until the bremsstrahlung endpoint reaches 12.438 MeV. Also, even when the bremsstrahlung endpoint reaches that value, neutrons from the (γ, n) process will be very small in number because they can only be produced by the few photons at the bremsstrahlung endpoint energy. These energetic considerations apply in a similar manner for all fissionable nuclear materials, in particular for those with Z≧89, the region of the actinides. In addition, and most importantly, most heavy elements such as Bi, Pb, W, Ta, etc. have isotopes with (γ, n) thresholds at or above 6.5 MeV. Therefore, finding neutrons with energies above E=Eb−Eth where Eth is in the range of 6 MeV constitutes a very definitive test for the presence of fissile material. Another test to verify that the detected neutrons result from photo-fission is the sensitivity of the yield of neutrons at energies above E=Eb−Eth to a modest increase in incident photon energy. In particular, measuring the increase in yield relative to the yield of neutrons below this energy is significant. The increase or relative increase in neutron yield is not substantial when the neutrons are emitted from photo-fission fission fragments because energetic considerations independent of the exact incident photon energy, such as the boost in velocity from fission fragment motion, are most important in determining the yield. FIG. 3 displays spectra of (γ, n) neutrons for gold. (It is FIG. 2 from W. Bertozzi, F. R. Paolini and C. P. Sargent, “Time-of-Flight Measurements of Photoneutron Energy Spectra”, Physical Review, 119, 790 (1958)). FIG. 3 illustrates how the nature of the (γ, n) process causes neutrons produced by that process to be concentrated mostly at low energies. The data in FIG. 3 are normalized to yield the same number of neutrons from the (γ, n) process in a reference target of 2D with neutron energies En>1.4 MeV. Because photon and neutron energy are uniquely related in the (γ, n) process in 2D, this normalization allows the formation of the difference photon spectrum (the difference between the high energy (15.8 MeV) bremsstrahlung spectrum and the low energy (14.3 MeV) bremsstrahlung spectrum), which corresponds to a broad band of photons centered at approximately 14.5 MeV and with approximately a 2 MeV half width. That is, the neutron energy spectrum produced by the difference in the neutron energy spectra at the two energies in FIG. 3 corresponds to photo neutrons produced by photons in the above energy band centered at approximately 14.5 MeV. FIG. 3 confirms that, because neutrons produced by the (γ, n) process are concentrated mostly at low energies, the contamination of a photo-fission spectrum by neutrons from the (γ, n) process is expected to be low at higher neutron energies, even when one looks at neutrons at energies below the E=Eb−Eth cutoff established by the strict application of energy conservation. The spectra in FIG. 3 show the rapid, almost exponential decrease of neutrons from the (γ, n) process with increasing neutron energy, in contrast to the neutron spectrum from the photo-fission (γ, f) of 232Th at 7.75 MeV bremsstrahlung energy as shown in FIG. 2. For gold the neutron spectrum from (γ, n) is nonexistent with if the bremsstrahlung spectrum endpoint is 7.75 MeV, since the (γ, n) threshold, Eth, is above 8 MeV. Even with a 12 MeV bremsstrahlung endpoint, the highest neutron energy from (7, n) in gold would be less than 4 MeV., and neutrons in this energy range from (γ, n) would not be numerous because they would correspond to photons at the endpoint of the bremsstrahlung spectrum. The neutron yield from (γ, f) in 232Th is very large at 12 MeV bremsstrahlung for neutron energies above 6 MeV. Table 1 gives the (γ, f) and the (γ, n) thresholds (in MeV) for some typical nuclei in the actinide region. The (γ, f) thresholds are from H. W. Koch, “Experimental Photo-Fission Thresholds in 235U, 238U, 233U, 239Pu and 232Th”, Physical Review, 77, 329-336 (1950). The (γ, n) threshold of 207Pb is also listed, as it is a component in natural lead material that may be used as a shield against detection of fissile materials. The table shows the maximum neutron energy available from the (γ, n) process for bremsstrahlung end point energies up to 11 MeV, including for 207Pb. This energy is to be compared to the spectrum in FIG. 3 showing many neutrons with energies in excess of 6 MeV from 232Th photo-fission using bremsstrahlung of only 7.75 MeV. Even with an 11 MeV bremsstrahlung energy there are no neutrons above 5.7 MeV from any nucleus, and no neutrons above 4.26 from 207Pb, and those at or near these energies would be very few in number because they correspond to the photons at or near the end-point energy of the bremsstrahlung spectrum. It should be noted that the (γ, f) process increases in importance as the bremsstrahlung endpoint energy increases from 6 to 11 MeV because of the increasing cross section with energy and because of the increasing number of photons in the bremsstrahlung spectrum at lower photon energies where the (γ, f) cross section is sizable. The (γ, f) thresholds are almost all lower than the (γ, n) thresholds, and are all significantly lower than the (γ, n) threshold for 207Pb. TABLE 1Maximum Neutron Energies from (γ, n) for SelectedBremsstrahlung Energies and Isotopes.Maximum (γ, n) Neutron Energy (MeV)(γ, f) Threshold(γ, n) ThresholdBremsstrahlung γ Endpoint Energy, Eb (MeV)Element(MeV)(MeV)67891011232Th5.40 ± 0.226.438—0.561.562.563.564.56233U5.18 ± 0.275.7590.241.242.243.344.345.34235U5.31 ± 0.275.2980.701.702.703.704.705.70238U5.08 ± 0.156.154—0.851.852.853.854.85239Pu5.31 ± 0.255.6470.351.352.353.354.355.35207Pb—6.738—0.261.262.263.264.26 The data in Table 1 indicates how the yield of neutrons above a specified energy would change as the bremsstrahlung endpoint energy is changed. For 207Pb, Table 1 indicates, there would be no neutron yield above 4 MeV until the electron beam energy exceeded approximately 11 MeV. (For gold, as discussed above in connection with FIG. 3, the electron beam energy would have to exceed 12 MeV to provide a neutron yield above 4 MeV.) However, the yield of neutrons above 4 MeV for the actinides would be a strongly increasing function of electron beam energy starting below 6 MeV electron beam energy since the low (γ, f) threshold allows the photo-fission process to grow rapidly as more and more photons are available for photo-fission, all of them producing a neutron spectrum independent of photon energy and strongly populating the selected region of neutron energy (above 4 MeV for example). The (γ, n) process in the actinide examples shown in Table 1 or in other heavy metals such as 207Pb would not be a significant component of the total yield until the electron beam energy is well above 10 MeV since the process involves only the photons near the bremsstrahlung endpoint, Eb. An additional point, which will be discussed further below, is that the photo-fission cross section is larger than the (γ, n) cross section over most photon energies by a considerable amount, as shown in FIGS. 5B and 5D. The neutrons from (γ, f) will dominate (γ, n) in most situations simply on the basis of the cross sections, aside from the other features discussed herein. The data in Table 1 is based upon continuous bremsstrahlung spectra with specific endpoint energies, but a similar discussion applies to monochromatic photon beams. The neutron energy spectra from photo-fission retains the same dependence on neutron energy for different photon energies, but the total yield is modulated for monochromatic photons only by the cross section for (γ, f) at the specific photon energy. In contrast, the total yield for neutron production from a bremsstrahlung beam is modulated by the convolution of the bremsstrahlung spectrum with the (γ, f) cross section. The maximum neutron energy from (γ, n) dictated by energy conservation considerations for monochromatic incident photons follows just as discussed above. Other energies than 4 MeV could be used as the “trigger” or cutoff for defining the presence of fissionable nuclear material. That is, for any specific electron beam energy, a “trigger” energy can be selected such that the presence of neutrons with an energy above that “trigger” energy will be energetically impossible for the (γ, n) process in relevant heavy materials such as 207Pb and therefore any neutrons detected could only originate from the photo-fission process in an actinide. The data in FIG. 2 show that there are many neutrons above 6 MeV from the (γ, f) process, and hence 6 MeV could be selected as a “trigger” energy. Other “trigger” energies are possible also; the choice is dependent on factors such as the speed of detection that is desirable, the false positives that are to be allowed, and the efficiency of detection that is desired. In addition, the choice may be dictated by the specific nature of the cargo in a container; if the cargo is made of materials with high (γ, n) thresholds, such as copper, aluminum, steel or oxygen, then a lower trigger could be selected. Conversely, hydrogenous material that naturally contains a small percentage of deuterium may be of concern because of its low threshold for the (γ, n) process, 2.2 MeV. However, because the energy release is shared almost equally by the neutron and proton, the maximum neutron energy is given by E=(Eb−2.2)/2 MeV and, for the example of an electron beam energy of 10 MeV, the maximum neutron energy is approximately 3.9 MeV and a 9.2 MeV photon results in a neutron energy of 3.5 MeV. Thus, a higher trigger may be appropriate A more important concern may be 9Be. It has a low (γ, n) threshold of only approximately 1.6 MeV and the energy sharing results in a neutron that has most of the available energy, E=(8/9)(Eb−1.6) MeV is the maximum neutron energy available. For the example of Eb=10 MeV, the maximum neutron energy is approximately 7.5 MeV. This high energy could present a serious background. However, one could distinguish neutrons from actinide photo-fission from neutrons from the (γ, n) process in 9Be by taking advantage of the fact that the (γ, n) process follows the strict rule for conservation of energy, so that E=(8/9)(Eb−1.6) defines the maximum neutron energy possible, while the photo-fission process has a neutron energy spectrum largely independent of the photon energy in the energy region under discussion, Eb less than approximately 15 MeV. Therefore, neutrons at an energy greater than E=(8/9)(Eb−1.6), where Eb is the photon beam energy or bremsstrahlung endpoint energy, is proof of a fissile material. At Eb=10 MeV, the presence of neutrons above approximately 7.5 MeV would be proof. At Eb=8 MeV, neutrons above 5.7 MeV would be proof. Also, the prompt neutron energy spectrum is independent of the photon energy while the (γ, n) process in 9Be produces a neutron spectrum that is strongly dependent on photon energy. This difference also permits distinguishing the presence of a fissionable element from the presence of 9Be. However, if there were concern that this measurement could not be reliably made, further steps could be taken. Operating at Eb=10 MeV, the maximum neutron energy from beryllium (γ, n) is approximately 7.5 MeV. By reducing the beam energy to 8 MeV, for example, the maximum energy neutron from beryllium (γ, n) would be reduced to 5.6 MeV but the photo-fission neutron energy distribution would be unchanged. If there are neutrons above 5.6 MeV the process is unquestionably photon induced fission. If there remains any doubt that neutrons are from fission, the photon beam energy can be further reduced. For example at 5 MeV photon or bremsstrahlung beam energy there will be little or no photo-fission. But beryllium (γ, n) will produce neutrons of up to approximately 3 MeV at that photon beam energy. The presence of these neutrons will clearly establish the presence of beryllium. From the yield of these neutrons, the contributions from beryllium to higher neutron energies when higher photon energies are used can be calculated, the neutron energy distribution from beryllium removed, and the remaining spectrum analyzed for the presence of actinide neutrons. Fortunately, 9Be is almost unique in this category. There are a few other nuclei with relatively low (γ, n) thresholds; 6Li, 13C, 17O and 149Sm are notable among these with thresholds of 5.66, 4.95, 4.14 and 5.87 MeV, respectively. The same procedures outlined above can be used to eliminate these sources as contributors masking fissionable nuclei. FIGS. 4A and 4B, from H. W. Koch, “Experimental Photo-Fission Thresholds in 235U, 238U, 233U, 239Pu and 232Th”, Physical Review, 77, 329-336 (1950), FIGS. 4 and 5, display the yield of fission fragments as a function of bremsstrahlung endpoint energy (“Peak Spectrum Energies”) for two isotopes, 235U (FIG. 4A) and 239Pu (FIG. 4B). These illustrate the rapid increase of the fission yield as a function of the energy of the electron beam used to produce bremsstrahlung. FIG. 4A also shows the dominance of the 235U contribution over the impurities of 238U in the enriched uranium sample. These data are based upon the detection of the actual photo-fission fragments. The yield of prompt neutrons follows approximately the same yield curve, since neutron emission in the photo-fission process is not dependent on the photon energy in the regions of interest below the Giant Electric Dipole Resonance at approximately 12 to 13 MeV photon energy. The emission of neutrons from the fragments is determined by the complex dynamics, discussed earlier, of splitting the fissioning nucleus into two fragments. As a result, the shape of the yield curve of prompt neutrons of a given energy as a function of bremsstrahlung energy will be essentially independent of the neutron energy. That is, the yield curve for 6 MeV neutrons will have the same dependence on bremsstrahlung endpoint energy as the yield curve for 2 MeV, 3 MeV, 4 MeV and etc. neutrons. This is in contrast with the yield curves for neutrons from the (γ, n) process, which will start at the endpoint energy given by Eb=Eth+En, where En is the neutron energy that is desired. They are thus displaced from the (γ, n) threshold energy, Eth, by the neutron energy, in contrast to the yield curves for (γ, f). This is a powerful signature that the neutrons detected are from photo-fission rather than from (γ, n). FIGS. 5A through 5D display the photon induced reaction cross sections for 239Pu. They are taken from FIG. 7 of B. L. Berman, J. T. Caldwell, E. J. Dowdy, S. S. Dietrich, P. Meyer, and R. A. Alvarez, “Photofission and photoneutron cross sections and photofission neutron multiplicities for 233U, 234U, 237Np and 239Pu”, Physical Review C Volume 34, Number 6, 2201-2214 (1986). FIG. 5A shows the total photon absorption cross section. FIG. 5B shows the partial cross section for (γ, 1n), single neutron emission. FIG. 5C shows the partial cross section for (γ,2n), double neutron emission. FIG. 5D shows the partial cross section for (γ, f), photo-fission. The photo-fission cross section (FIG. 5D) is larger than the (γ, n) cross section (FIG. 5B) over most photon energies by a considerable amount. This displays the feature common to the actinides that photo-fission is a strong and often dominant process from the (γ, f) threshold throughout much of the Giant Electric Dipole Resonance. Given that the prompt neutron multiplicities from photo-fission range from approximately 2.5 to more than 3, prompt neutrons from the photo-fission process will dominate the incident photon reaction channel by a large factor at all neutron energies. This feature facilitates identifying the presence of actinide fissionable material despite the potential presence of other heavy elements such as Pb, even without considering the energy-conservation constraints on neutron energy. The photon absorption process in most heavier nuclei is dominated by neutron emission, and the total yield is governed by the giant dipole sum rule that the integrated cross section is proportional to NZ/A, which is a slowly varying function. Since the location of the giant dipole resonance in energy also is a slowly varying function of nuclear mass, a yield of prompt neutrons from the photo-fission process that is 2.5 to 3 times the expected neutron yield from (γ, n) is a signal of photo-fission in that the (γ, f) neutron yield alone will produce a markedly higher photon absorption cross section than would (γ, n) for a given quantity of heavy material. That is, measuring the yield of neutrons per heavy nucleus per photon permits identifying photo-fission as present, if the quantity of heavy material present can be determined by measuring localized density by other methods. The angular distribution of the prompt neutrons and the relationship of the neutron energy to the fragment angular distribution also are signatures of fissile material and the photofission process, and can be used in detection schemes. The fragment angular distributions are not as distinct for odd-even nuclei as for even-even nuclei, in part because of the high population of spin states. Odd-even nuclei angular distributions are almost isotropic as reported by L. P. Geraldo, “Angular Distribution of the Photofission Fragments of 237Np at Threshold Energy”, Journal of Physics G: Nuclear Physics, 12 1423-1431 (1986), which shows angular anisotropy of approximately 10% at 5.6 MeV, 6% at 6.61 MeV and 2% at 8.61 MeV. These results are very much in contrast with the large anisotropy for fragments from the photo-fission of even-even nuclei where ground state spins are zero. Thus, once actinide photo-fission is detected, a nearly isotropic neutron angular distribution is an indicator of an odd-even fissile species such as 235U, 237Np and 239Pu. A strongly anisotropic neutron angular distribution would indicate an even-even fissile species such as 232Th and 238U. (See S. Nair, D. B. Gayther, B. H. Patrick and E. M. Bowey, Journal of Physics, G: Nuclear Physics, Vol. 3, No. 7, 1977 (pp 1965-1978) and references therein, for example.) The energy distributions of the neutrons at various angles are themselves indicators of the fragment anisotropy, and thus of the type of nucleus. This fact was used in the analysis of the work by Sargent et al, discussed above. If the fragments are strongly anisotropic (even-even fissile species), then the energy spectra of the neutrons will show distinct differences at different directions with respect to the photon beam. As an example, if the fragments are strongly peaked at 90 degrees with respect to the photon beam, then the neutron spectrum at 90 degrees will exhibit to a different degree the boost in velocity due to the velocity of the fragments than the neutron spectrum at angles near 180 degrees or 0 degrees to the photon beam. However, if the fragment angular distribution is nearly isotropic (odd-even fissile species), then the energy distribution of the neutrons will be the same at all angles. In both situations, the higher energies reflect the motion of the fragments, but the contrast in the energy distribution of the neutrons at different angles will reflect the fragment anisotropy with angle. The fragment angular distributions dominate the neutron angular distributions and the neutron energy distributions as a function of angle. The results of E. J. Winhold, P. T. Demos and I. Halpern, Physical Review, 87, 1139 (1952); E. J. Winhold and I. Halpern, Physical Review, 103, 990-1000 (1956); and, A. P. Berg, R. M. Bartholomew, F. Brown, L. Katz and S. B. Kowalski, Canadian Journal of Physics, 37, 1418 (1959) show the fragment angular distributions for various isotopes. The following abstract from Berg et al. is offered as a summary of the data in that paper: Angular distributions of photofission fragments relative to the photon beam have been measured as a function of maximum bremsstrahlung energy in the range 6-20 MeV. The nuclides U-233, U-235, Np-237, Pu-239 and Am-241 give an isotropic distribution at all energies studied. The nuclides Th-232, U-234, U-236, U-238, and Pu-240 give anisotropic distributions which can be described by an equation of the form W(Θ)=1+α sin2 Θ, where Θ is the angle between fragment and beam. The degree of anisotropy is large at low energy and falls rapidly as the energy is increased. At a given energy Th-232 has the greatest degree of anisotropy and Pu-240 the least. The result quoted in the abstract is in basic agreement with that of the other papers referred to herein. In addition, some greater detail about the results from Berg et al. is shown in the two tables taken from that reference: TABLE 2Angular Distributions (from Berg, et al. Table I)Angular distributionsRatio, counts at 90°/counts at 0°NuclideE0† = 6.0E0 = 6.5E0 = 8.0E0 = 10.0E0 = 20.0U-2331.048 ± 0.071.032 ± 0.040.994 ± 0.03U-2351.024 ± 0.05Np-2371.024 ± 0.10Pu-239‡1.034 ± 0.260.927 ± 0.121.002 ± 0.061.013 ± 0.050.952 ± 0.03Am-2410.958 ± 0.07The ratio is the number of counts observed at 90° per unit X-ray dose divided by the number observed at 0° for the same dose.†E0 is the maximum energy in million electron volts of the bremsstrahlung spectrum.‡The 45°/0° ratio at E0 = 6.5 Mev was 1.09 ± 0.23. TABLE 3Corrected Values of α (from Berg, et al. Table VI)Corrected values of α in W(θ) = 1 + α sin2 θE0Th-232U-238U-236U-234Pu-2406.06.6 ± 2 6.0 ± 2.36.36.7 ± 1.16.5>254.4 ± 1.02.1 ± 0.42.3 ± 0.60.65 ± 0.207.011.0 ± 0.8 2.05 ± 0.241.33 ± 0.170.90 ± 0.160.49 ± 0.127.510.3 ± 1.6 8.04.9 ± 0.61.3 ± 0.10.79 ± 0.090.44 ± 0.080.29 ± 0.079.02.8 ± 0.40.51 ± 0.079.40.44 ± 0.0410.01.61 ± 0.120.41 ± 0.050.32 ± 0.060.17 ± 0.0714.00.46 ± 0.090.09 ± 0.040.04 ± 0.0315.0 0.02 ± 0.04* 0.01 ± 0.0320.00.14 ± 0.060.05 ± 0.03These values, which do not differ from zero, have not been corrected for isotopic composition. Table 2 (“Angular Distributions . . . ”) shows that the ratio of events at 90 degrees to those at 0 degrees for the photo-fission of the odd-even isotopes shown is approximately equal to 1 over the energy range of the bremsstrahlung endpoints shown in the table. Thus, the value of b/a discussed earlier is 0 and the angular distribution is isotropic. Table 3 (“Corrected values . . . ”) shows the fit to the normalized form of the angular distribution as exhibited in the table also as a function of bremsstrahlung endpoint. The derived angular distributions are clearly anisotropic. From these data, the quoted abstract, and the theoretical basis referred to in the references herein, the generalization is accurate; the odd-even actinides undergo isotropic photo-fission while the even-even actinides undergo anisotropic photo-fission. In particular, the result is experimentally demonstrated for the isotopes most likely to be used for a nuclear weapon, 235U, 239Pu and 237Np. These will undergo isotropic photo-fission, in contrast to 238U, 232Th and the other even-even isotopes that were measured. FIG. 10, which displays the fission fragment angular distributions from photo-fission of an even-even nucleus, and FIG. 7, which displays the angular distributions of the prompt neutrons emitted from those fragments, demonstrate the general peaking of the fragment and neutron angular distributions at 90 degrees relative to the photon beam. The neutron yield at 150 degrees is approximately 20% less than that at 90 degrees and the shape of the distribution is approximately symmetric about 90 degrees. The fragment angular distributions show a larger anisotropy as expected because the neutron distributions are produced by folding the isotropic angular distributions in the fragment center of mass with the fragment distributions in angle. In contrast to these distributions, the isotopes 235U, 239Pu and 237Np (not shown), which may be used in the manufacture of weapons, display for the most part isotropic angular distributions of the photo-fission fragments as discussed above and the resulting angular distributions of the prompt neutrons also are isotropic. Having reviewed the case of photon-induced fission in detail, we now turn to neutron-induced fission, the subject of this disclosure. As discussed earlier, the energy distribution of prompt neutrons emitted from neutron induced fission, photo-fission and spontaneous fission are very much alike and depend more on the specific nucleus undergoing fission than the energy of the initiating photon or neutron. These remarks pertain to fission initiating neutron energies of less than approximately 15 MeV and photon energies in or below the Giant Dipole region. The prompt neutrons from fission appear to be emitted for the most part from the fission fragments that have achieved their fully accelerated velocities. In the frame of reference of these fragments the prompt neutrons are emitted isotropically in angle with an energy distribution that is characterized by the energetic aspects of the fission process for the specific fissionable nucleus and is “thermal” in character. As a result of the boost from the fragment velocity, the neutron angular distributions also reflect the angular distributions of the fragments relative to direction of the initiating particle. In addition, the energy distribution of the prompt neutrons is also correlated to the angle of the prompt neutrons when the fission fragments are not emitted isotropically relative to direction of the initiating particle. These effects were demonstrated explicitly for photo-fission by the data of Sargent et al. and S. Nair, D. B. Gaythwer, B. H. Patrick and E. M. Bowey, Journal of Physics, G: Nuclear Physics, Vol. 3, No. 7, 1977 (pp 1965-1978). They are also demonstrated for neutron induced fission by the data of S. Nair and D. B. Gayther, Journal of Physics G: Nuclear Physics, Vol. 3, No. 7, 1977 (pp 949-964). The angular distribution of the fission fragments in neutron induced fission plays a dominant role in determining the angular distribution of the prompt neutrons that are emitted. For fission induced by un-polarized neutrons of low energy the fragments are emitted isotropically since only J=½ (angular momentum) is involved in the reaction process. As the energy of the initiating neutron increases the fragments can display anisotropic angular distributions. This is demonstrated by the work of S. Nair and D. B. Gayther, Journal of Physics G: Nuclear Physics, Vol. 3, No. 7, 1977 (pp 949-964) and also by other authors such as the data of Henkel and Brolley used by E. Hyde in UCRL 9065 and illustrated in FIG. 13 reproduced from “NUCLEI AND PARTICLES An Introduction to Nuclear and Subnuclear Physics”, Emilio Segre, W. A. Benjamin, Inc., New York 1964 (page 502). These data show that the fragment distributions from neutron induced fission become peaked forward and backwards at the higher energies as a result of the M=0 dominance of the orbital angular momentum projection relative to the neutron direction. At energies below ˜3 MeV the angular distribution of the fragments varies with energy in a more irregular manner. Nevertheless, it is possible to calibrate these dependences accurately for each fissile isotope and the fragment angular distributions of prompt neutrons resulting from fast neutron induced fission can be used to specify further the differences between fissile isotopes. In fact, the prompt neutron angular distribution and its dependence on prompt neutron energy can be calibrated explicitly for different initiating neutron energies without reference to the fragment distributions. These differences in prompt neutron angular distributions and energy distributions correlated to angle can be used to distinguish different fissile isotopes. Equally important in this disclosure is the fact that the energy of the initiating neutron can be controlled. For example the D(D,n)3He reaction will produce neutron energies of approximately 2.5 MeV for modest deuteron energies of 50 to 200 KeV. Using such a source would allow the detection of a fissile material simply by detecting neutrons from a container above 2.5 MeV. The neutron source would not be an important source of background. Other reactions can be used to generate neutrons of a defined energy or range of energies. Among others, these include the 3H(D,n)4He, (p,n), (α,n) reactions and those involving the use of radioactive materials that emit neutrons or alpha particles used to bombard beryllium to produce neutrons. Some of these reactions and sources produce neutrons of an energy that is higher than a few MeV and the 3H(D,n)4He reaction is one of these with neutrons of approximately 14 MeV. The neutron source in these cases must be carefully shielded and collimated so as to not be mistaken as a fissile material because of the high energy neutrons contained in the spectrum. One technique that is used to control the energy of such neutron sources is to scatter these neutrons from a light nucleus such as hydrogen and by means of collimation and shielding only allow the neutrons scattered through a specific angle to impinge on the container being interrogated. This angle determines the final neutron energy and for example scattering from hydrogen at 45 degrees will reduce the neutron energy from its original energy E to E/2. Other such scattering situations will be understood by those skilled in the art and are included in this disclosure. One embodiment of a detector system to carry out the methods described hereinabove with respect to photon-induced fission is described in U.S. patent application Ser. No. 12/139,050, to which this application claims priority, and which is incorporated herein. That embodiment requires a source of photons with energy capable of exceeding the (γ, f) threshold and a detector for neutrons. The photons may be monochromatic, may be produced by a source capable of variable energy, or may be distributed over a broad range of energy with a good definition of the highest energy possible, such as an electron-generated bremsstrahlung spectrum in accordance with the discussion above. When an accelerator is used to provide the electrons, the electron accelerator may have the capability to vary the energy of the electron beam from below the fission barrier (threshold) to higher energies in order to exploit all the modalities discussed above. Any neutron detector that is capable of distinguishing neutron energy is appropriate. A detector that takes advantage of energy deposition, such as proton recoil from neutron elastic scattering in a hydrogenous scintillator, is a possible choice. A detector that measures a reaction energy induced by the neutron is another possible choice. A method of measuring neutron energy by time of flight is also an appropriate detection scheme. The energy resolution required for such detection methods will have to be sufficient to eliminate neutrons from the (γ, n) process in materials other than actinides, as discussed above. Because the contamination of non-actinide (γ, n) can be controlled and rendered small by the choice of incident photon energy (or bremsstrahlung endpoint) and neutron energy measured, the resolution required is well within a number of measurement techniques. Specific resolutions required may depend in detail on the particular situation under consideration, but resolutions of approximately 0.5-0.75 MeV at 4 to 6 MeV neutron energy may be adequate. A detection method may be required to operate in a possible flux of photons in some embodiments, these photons being produced by scattering from the material under study in the direction of the detectors. Photons may also be produced by natural radioactivity and cosmic rays. Therefore, the neutron detectors may have to be shielded using passive and active shielding techniques. In addition, as a consequence of the above, a neutron detector may be required to distinguish between photons and neutrons. This can be accommodated by the reaction process used, the time of flight of the photons compared to neutrons and by the ability of the detector to distinguish between the deposition of energy by heavy particles (e.g., neutrons) compared to electrons. Organic and inorganic scintillators that have different decay times according to the density of ionization produced by the passage of a charged particle may be suitable. Separation of photons from neutrons may be achieved in such scintillators utilizing signal processing techniques that exploit these different charged particle responses. FIG. 12 demonstrates the energy separation of prompt neutrons from photon induced fission from neutrons produced from the (γ,n) reaction, when incident photon energies below 9 MeV are used. That Figure was obtained using 9 MeV bremsstrahlung beams produced at the CW S-DALINAC at the Technical University of Darmstadt. The spectra from Pb and highly enriched uranium (HEU) targets shown in FIG. 12 were obtained using the technique proposed for the CAARS PNPF module. An organic liquid scintillator was used to determine the energy deposited in the detector and to separate photon and neutron events. The lighter data points represent events from Pb (which as discussed above has a (γ,n) threshold of approximately 6.5 MeV for the 207Pb isotope) and the darker represent events from HEU (which has a photo-fission threshold of 5.5 MeV). The neutron events are grouped in the lower portion of the figure and are clearly differentiated from the photon events above. Within the neutron events, the box shows where the neutrons from prompt photo-fission in HEU appear unambiguously. A neutron signal in this region is an unambiguous signal of an actinide photo-fission event. One exemplary embodiment of a system 600 for detecting fissile materials in a container by analyzing energetic prompt neutrons resulting from photon-induced fission is illustrated in FIG. 6. An electron beam 602 of energy Eb is generated by an electron accelerator 601. The electron beam 602 makes bremsstrahlung radiation photons when it strikes a bremsstrahlung target 603 (also called the radiator). The electron accelerator 601 and radiator 603 optionally may be replaced by a source of monochromatic or nearly monochromatic photons. The optional collimator 604 collimates the bremsstrahlung radiation. A shield 605 may enclose the bremsstrahlung target 603 and electron accelerator 601. The photon beam 607 is directed onto a container 606 which is to be analyzed and which may contain fissile material 608. The distances of the fissile material 608 from (for example) three of the walls of container 606 are designated as x, y and z. A photon detector 609 placed after the container 606 optionally may be used to monitor the transmitted photon flux of photon beam 607. Detectors 610, 611, 612, and 613 may be placed at locations around the container 606 at approximately 90 degrees and at convenient back angles with respect to the collimated photon beam 607. The number and location of the detectors may be varied from that shown in FIG. 6 according to the principles and methods discussed above. In the illustrated embodiment, the detectors 610 and 611 may be placed at known distances L610 and L611 from the container 606 walls. The detectors 610, 611, 612, and 613 optionally may be surrounded by shielding (not shown) and by anti-coincidence counting systems (not shown) if desired. The detectors 610, 611, 612, and 613 themselves may be sensitive to neutron energy or they may be part of a system (such as one utilizing time-of-flight) that will provide a neutron energy for each detected neutron event. A beam dump 614 may be used to absorb the remaining photon flux after the photon beam 607 passes through the container 606 and its contents. The beam dump 614 and optional transmitted flux monitor, detector 609, may be shielded from direct view of the detectors as required. Signals from the detectors 609, 610, 611, 612, and 613 are connected by way of connections 615 to signal processing electronics and/or computer 616, which process the detector signals and optionally may relay them and/or processed information by way of connections 617 to a central control and data analysis and storage system (not shown.) Alternatively, the detector signals may be passed directly to the central system for processing and analysis. As an alternative to determining neutron energy directly in the neutron detector, a low duty cycle LINAC (e.g. Varian linatron) or other suitable electron accelerator may be pulsed to permit a time of flight (TOF) technique. Compared to other detection techniques, such as pulse shape discrimination using a continuous incident photon beam, the TOF method is expected to have a higher efficiency for collecting high energy neutrons, reduced environmental background, and a higher likelihood of determining angular distributions. The TOF method may use a shortened pulse structure (10 ns) and gated detectors to reject gamma flash. The advantages inherent in the TOF method, combined with the modified LINAC and detectors, may partially compensate for the reduced duty cycle of commonly deployed pulsed accelerators. In a time-of-flight (TOF) embodiment, the electron accelerator 601 or other source may be pulsed to produce electron beam 602 (pulsed on) for a time period T and turned off for a time long enough to have all the detectable neutrons (resulting from interactions of the photon beam 607 with the container 606 and its contents) pass through the detector(s). Then the electron beam 602 may be pulsed on again for a time period T. This sequence may be repeated until the desired detection data is obtained. The electron accelerator 601 or some subsidiary target (not shown) near the bremsstrahlung target 603 or in the bremsstrahlung or photon beam 607 may provide a fiducial signal that informs the signal processing electronics and/or computer 616 when the photon beam 607 was generated. Neutrons generated by photofission in the fissile sample 608 travel to a detector in the time L/v where L is the distance from the fissile sample 608 to the detector in question and v is the neutron velocity. For detector 611, for example, which is opposite the fissile sample 608 at a right angle to the incident photon beam 607 in the embodiment shown, L=L611+y, the distance from the fissile sample 608 to the corresponding wall of the container 606 nearest detector 611. The velocity of the neutrons is given by v=(2E/m)1/2, where E is the neutron kinetic energy and m is the neutron mass. The signal from detector 611 goes to the signal processing electronics and/or computer 616, which converts the difference between the fiducial signal arrival time and the detector 611 signal arrival time into the time-of-flight (TOF) of the neutron to the detector. Using the relation TOF=(L611+y)/v, the signal processing electronics and/or computer 616 calculates the neutron velocity and therefore its energy (E=mv2/2) and records the data and also transfers it to a central control and analysis system (not shown). The energy resolution of the detection system will depend on the TOF of the neutrons, T, L and the dispersion of the flight distance to different portions of the detectors. Those experienced in the art will recognize that these parameters, including the electron beam pulse width T, and the geometry of the system can be adjusted to achieve energy resolution adequate for the purposes of this disclosure. The (narrow) photon beam 607 may be scanned across the container 606 sequentially to illuminate discrete columns where the fissile sample 608 may be located. This serves to better localize the position of any fissionable material and will reduce backgrounds from other neutron producing materials in a container. Alternatively, the photon beam 607 may be a wide fan-like beam encompassing a greater region of the container 606 with the fan opening out in the direction toward the detectors at 90 degrees, for example. This allows a broad scan region of the container but limited in the narrow direction. Such an embodiment would facilitate scanning the container in shorter times for fissile materials. It would detect fissile materials distributed over the dimensions of the fan beam. In this geometry x and y will not be known but they may be inferred from a comparison of the neutron energy spectra on both sides of the container since they should be very close to identical, especially at the highest energies. Starting with any assumption for “a”, such as ½ the width of the container (x=y), the resulting spectra can be adjusted by varying “a” until the spectra are made to have the same high-energy shape. The technology for short duration electron beam pulses is a well-known art, and pulses of a few nanoseconds are readily generated for high energy electron beams. Time of flight for a 1 MeV neutron over 1 m is 72 nanoseconds. Thus, flight distances of a few meters result in flight times (˜71 nanoseconds for 6 MeV neutrons over a distance of 3 meters, for example) that allow beam pulse duration times of 10 to 20 nanoseconds to separate photo-fission neutrons from those from (γ, n) processes by energy selection. Other specific embodiments are possible and some are mentioned herein as further illustrations of methods to articulate the concepts and methods described earlier. The detectors 610, 611, 612, and 613 in FIG. 6 can be any that are capable of unambiguously detecting a neutron. Rather than measuring the neutron time of flight to determine its energy, it would suffice in some applications to only specify that the event is definitely a neutron and that the energy is greater than a defined amount. This would characterize the neutron energy as above a defined quantity. Several such neutron energies may be involved. Together with control of the electron beam energy or photon energy as discussed above, determining the number of neutrons with energies above certain preset quantities will classify the neutrons as from photo-fission. As discussed above, other processes such as (γ, n) will not be possible at neutron energies greater than E=Eb−Eth, where Eb is the bremsstrahlung endpoint or the photon energy and Eth is the threshold for (γ, n) for relevant non-actinide materials that may be present and need to be distinguished from the suspected actinide. As discussed above, the energy distribution of neutrons from photo-fission is very independent of the energy of the photons used to induce photo-fission in the photon energy regions discussed herein, in or below the Giant Electric Dipole Resonance. Another embodiment uses this fact to determine whether the neutrons originate from photo-fission. Varying the photon energy or the bremsstrahlung endpoint energy will not substantially alter the energy distribution of the neutrons from photo-fission. However, this is not true for other processes such as (γ, n), especially in the higher regions of neutron energy, as a result of energy conservation and the requirement E=Eb−Eth discussed earlier. Therefore, measuring the energy distribution of the neutrons for different photon energies, and comparing the results, can identify actinide photo-fission. Alternatively, measuring and comparing the number of neutrons above a certain energy as the photon energy is changed can achieve the same result. Another embodiment would measure the neutron yield at a given neutron energy, as the photon energy is varied, and would do this for several neutron energies. This would generate yield curves for neutrons of the given energies as a function of photon energy. Because the neutron energy spectra from photon-induced fission is independent of the incident photon energy, the same yield curve as a function of photon energy would result for all neutron energies if the spectrum is dominated by photo-fission. However, if the neutron spectrum originates from (γ, n) for relevant non-actinide materials, each neutron energy has a yield curve as a function of photon energy displaced in photon energy by that explicit neutron separation energy, in particular for the neutrons at the highest energy possible. Once again this follows from energy conservation. Neutron detection can be based on reaction energies between the neutrons and the component materials in the detector. Detectors of such a nature may sometimes but not always be called “threshold detectors” because a reaction will occur only if the neutron energy is greater than a certain amount. Examples of such reactions include but are not limited to (n, n′γ), (n, n′f), (n, n′p), (n, n′d) and (n, n′α). Detection of the event may be based on, but not limited to, the detection of: a scintillation event and measuring the deposited energy; the charge created by ionization in a material and measuring the total charge; and, the detection of radioactive nuclei, wherein the radioactivity would be induced only if the neutron energy (energies) were greater than a certain value (or values). All such methods are included in the embodiments described in this disclosure. As discussed above, some commercially available plastic and liquid scintillators can identify neutrons unambiguously using suitable signal processing techniques. Such detectors also have fast enough time response to qualify for the purposes herein and these will be known to those skilled in the art. Such detectors operate in part as proton recoil detectors, based on the energy imparted to protons by the elastic scattering of neutrons from the protons in the hydrogenous material. Therefore, in part, they can function as “threshold detectors” as discussed above, as well as providing the time for an event in a detector and identifying the event as a neutron. Such detection methods are part of the embodiments described herein. Delayed neutrons following beta decay can also be detected by the methods discussed herein and serve as a method of detecting fissile materials. They will be less abundant than prompt neutrons by a very large factor, as discussed above. In most cases their presence can be used as a further detection method to augment the embodiments discussed herein. They can be distinguished from prompt neutrons by several techniques. Using TOF with a pulsed beam set to measure prompt neutrons, delayed neutrons appear as a uniform distribution in time that builds up with exposure time or the number of pulses in the TOF embodiment discussed above. The time for buildup of the delayed neutron signal is characteristic of beta-decay lifetimes. If the beam is turned off they will diminish in times characteristic of beta-decay lifetimes. The presence of the delayed neutrons may be neglected in many situations as a minor contribution. In some cases they may be used as an aid to the detection of fissile material. In all situations, the presence of delayed neutrons may be accounted for and the results corrected accordingly if the correction is required by these embodiments. The photon beams may be of the pulsed variety described above in discussing TOF embodiments, or they may be of continuous character as from continuous duty radiofrequency accelerators, DC accelerators or similarly functioning photon sources of a monochromatic or nearly monochromatic nature. Another scan embodiment would employ a very broad beam geometry in all directions transverse to the beam direction with collimation so as to limit the beam size to that of the container width in its largest manifestation. This embodiment would be very effective in the detection of fissile materials dispersed in small samples over a large volume, such as thin sheets broadly distributed over a large region of the container or small pellets broadly distributed. Many beam geometries are possible, each with specific advantages for certain situations as will be recognized by those skilled in the art, and they are all included in this disclosure. In order to carry out scanning of containers as rapidly as possible, it may be preferable to carry out an initial scan with a low threshold or trigger neutron detection energy, in order to maximize the signal from photofission, even at the cost of obtaining a signal from (γ, n) processes. If no events are recorded from the container or a portion thereof in an appropriate interval, or no events above an acceptable background, the scan can be continued to a further portion of the container, or the container can be passed on if the entire container has been scanned. If events are detected, the threshold or trigger neutron detection energy can be increased, and the container or portion thereof rescanned, using the higher neutron threshold or trigger detection energy to reduce or eliminate the contamination from the competing (γ, n) processes. Alternatively, of course, other of the methods set forth herein for discriminating between photofission and (γ, n) processes can be employed in the rescan. Because angular distributions may be difficult to measure given the differential absorption and scattering of different cargo loadings, it is important to recognize that, as discussed above, if the energy distribution of the prompt neutrons is independent of angle relative to the photon beam, then the fragments are emitted isotropically and the fissile material is an odd-even isotope: however, if the prompt neutrons have a spectrum with greater population at the higher energies at 90 degrees to the photon beam relative to the prompt neutron spectrum at large angles near 180 degrees, then the fragments have an angular distribution peaking at 90 degrees and the fissile material is an even-even isotope. Therefore, measuring the neutron energy distribution at two angles will enable this determination to be made. Another embodiment removes the uncertainty in the energy distribution and angular dependencies of the prompt neutrons caused by the differential absorption along different paths that neutrons take in traversing a container to the different detectors. This embodiment directs the photon beam into the container in different directions. For example, in one arrangement the photon beam may enter the container from the top and the neutron detectors view the neutrons at 100 degrees to the beam and at 170 degrees from the beam. By altering the photon beam direction to enter from the side of the container the detectors change roles. That one previously at 100 degrees is now at 170 degrees and the one previously at 170 degrees is now at 100 degrees. However, the differential aspects of neutron absorption remain exactly the same. The two measurements now provide a clear indication of the influence on the neutron energy distribution of the angle of emission of the neutron relative to the photon direction as well as the angular distribution of the neutrons relative to the photon beam direction. As one particular feature, if the photo-fission process is isotropic the relative neutron yields in the detectors will not change. A change indicates anisotropy in the original photo-fission process. This process can be generalized for other angles as well. For example, FIGS. 8A and 8B, each show a container 806 with fissile material 808. A first neutron detector 801 is shown in a first location and a second neutron detector 802 is shown in a second location. In FIG. 8A, a photon beam 807A irradiates the container 806 from a first direction (direction 1). In FIG. 8B, a photon beam 807B irradiates the container 806 from a second direction (direction 2) For each of the two photon beam directions, the neutron detectors 801 and 802 interchange angles relative to the photon beam (807A or 807B) direction. For beam direction 1, first neutron detector 801 is at angle θ1 and second neutron detector 802 is at angle θ2. For beam direction 2, first neutron detector 801 is at angle θ2 and second neutron detector 802 is at angle θ1. I1, and I2 are the photon beam intensities at the target 808 (which may be a fissile material) for photon beam directions 1 and 2 respectively. If S(E,θ) is the energy spectrum of neutrons produced in direction θ, the neutrons detected by the two detectors with photon beam direction 1 are described by the measured functions Fi(E, θj): F1(E,θ1)=Ii×A1(E)×S(E,θ1) for first neutron detector 801; and, F2(E,θ2)=I1×A2(E)×S(E,θ2) for second neutron detector 802. The neutrons detected by the two detectors with beam in direction 2 are: F1(E,θ2)=I2×A1(E)×S(E,θ2) for first neutron detector 801; and, F2(E,θ1)=I2×A2(E)×S(E,θ1) for second neutron detector 802. The attenuation factors A1 and A2 remain invariant to the beam position and the ratio can be formed to eliminate these factors so that:{S(E,θ1)/S(E,θ2)}2={F1(E,θ1)×F2(E,θ1)}/{F2(E,θ2)×F1(E,θ2)}. (Equation 1) Thus, S(E,θ1) and S(E,θ2) are related via measured quantities and can be compared directly. A person skilled in the art will be able generalize this technique to more than two detectors and this embodiment is intended to contain all these variations This technique, wherein the absorptive properties of the material in the container may be removed from consideration, has applicability for both photon-induced and neutron-induced fission. Although the methods and systems discussed above have been described relative to photon-induced fission, they may be adapted and modified for application to neutron-induced fission. This is particularly true since the dynamics of these fission processes are regulated by the fact that these actinide nuclei are deformed with a low energy barrier to fission. Therefore, initiating fission only requires a small perturbation from the ground state to overcome the fission barrier. Given that the ground state is essentially a meta-stable state as evidenced by the existence of spontaneous fission, the details of the perturbation resulting in fission are not important to many properties of the final state. Hence, many of the details of fission depend most strongly on the nature of the fissioning nucleus, not the initiating particle or perturbation. The one strong exception to this ideal is the angular distribution of the fragments relative to the direction of the initiating particle. For the energy range discussed herein photon induced fission produces fragments that favor emission at 90 degrees to the photon beam, in particular for even-even actinides as discussed earlier. In contrast, neutron induced fission has a different behavior. Thermal neutrons induce fission that is largely isotropic with respect to fragment emission. As the neutron energy increase, we have shown above that the fragments in neutron induced fission can be emitted preferentially at 90 degrees and also forward and backwards with respect to the neutron beam. These behaviors reflect the importance of angular momentum preferences in the fissioning nuclei for neutrons of approximately 1 to 3 MeV while at higher energies the dominance of M=0 states in orbital angular momenta of the initiating neutron play a dominant role. Thus, with consideration of the details of the fission fragment angular distributions embodiments for photon induced fission may be adapted to apply to neutron induced fission with the replacement of the photon beam by a neutron beam. One exemplary embodiment of a system 900 for detecting fissile materials in a container by analyzing energetic prompt neutrons resulting from neutron-induced fission is illustrated in FIG. 9. A neutron beam 902 of energy Eb is generated by a neutron generator 901. The neutron beam may pass through a target (neutron filter) 903 to eliminate in some situations thermal neutrons generated within the shielding enclosure 905. The neutron generator may be a source of monochromatic or nearly monochromatic neutrons or may be replaced by such a source using radioactive materials. The optional collimator 904 collimates the neutron beam. A shield 905 may enclose the target 903 and neutron generator 901. The collimated neutron beam 907 is directed onto a container 906 which is to be analyzed and which may contain fissile material 908. The distances of the fissile material 908 from (for example) three of the walls of container 906 are designated as x, y and z. A detector 909 placed after the container 906 optionally may be used to monitor the transmitted neutron flux of neutron beam 907. Detectors 910, 911, 912, 913 may be placed at locations around the container 906 at approximately 90 degrees and at convenient back angles with respect to the collimated neutron beam 907. The number and location of the detectors may be varied from that shown in FIG. 9 according to the principles and methods discussed above. In the illustrated embodiment, the detectors 910 and 911 may be placed at known distances L910 and L911 from the container 906 walls. The detectors 910, 911, 912, 913 optionally may be surrounded by shielding (not shown) and by anti-coincidence counting systems (not shown) if desired. The detectors 910, 911, 912, 913 themselves may be sensitive to neutron energy or a minimum neutron energy or they may be part of a system (such as one utilizing time-of-flight) that will provide a neutron energy for each detected neutron event. A beam dump 914 may be used to absorb the remaining neutron flux after the neutron beam 907 passes through the container 906 and its contents. The beam dump 914 and optional transmitted flux monitor, detector 909, may be shielded from direct view of the detectors as required. Signals from the detectors 909, 910, 911, 912 and 913 are connected by way of connections 915 to signal processing electronics and/or computer 916, which process the detector signals and optionally may relay them and/or processed information by way of connections 917 to a central control and data analysis and storage system (not shown.) Alternatively, the detector signals may be passed directly to the central system for processing and analysis. In an embodiment the neutron beam 902 is not collimated to a very narrow cone and irradiates all or a substantial region of the container. In another embodiment, the neutron beam 902 initially may irradiate all or a substantial region of the container, but may be collimated to a narrow beam to localize detected fissionable material if the initial analysis with the broader beam indicates the presence of nuclear material in the container. In a time-of-flight (TOF) embodiment, the neutron generator 901 is pulsed to produce neutron beam 902 (pulsed on) for a time period T and turned off for a time long enough to have all the detectable neutrons (resulting from interactions of the incident neutron beam 907 with the container 906 and its contents) pass through the detector(s). Then the neutron beam 902 may be pulsed on again for a time period T. This sequence may be repeated until the desired data is obtained. The neutron generator 901 or some subsidiary target (not shown) near the neutron filter 903 or in the neutron beam 907 may provide a fiducial signal that informs the signal processing electronics and/or computer 916 when the neutron beam 907 was generated. Neutrons generated by neutron induced fission in the fissile sample 908 travel to a detector in the time L/v where L is the distance from the fissile sample 908 to the detector in question and v is the neutron velocity. For detector 911, for example, which is opposite the fissile sample 908 at a right angle to the incident neutron beam 907 in the embodiment shown, L=L911+y, the distance from the fissile sample 908 to the corresponding wall of the container 906 nearest detector 911. The velocity of the neutrons is given by v=(2E/m)1/2, where E is the neutron kinetic energy and m is the neutron mass. The signal from detector 911 goes to the signal processing electronics and/or computer 916, which converts the difference between the fiducial signal arrival time and the detector 911 signal arrival time into the time-of-flight (TOF) of the neutron to the detector. Using the relation TOF=(L911+y)/v, the signal processing electronics and/or computer 916 calculates the neutron velocity and its energy (E=mv2/2) and records the data and also transfers it to a central control and analysis system (not shown). The energy resolution of the detection system will depend on the TOF of the neutrons, T, L and the dispersion of the flight distance to different portions of the detectors. Those experienced in the art will recognize that these parameters, including the neutron beam pulse width T, and the geometry of the system can be adjusted to achieve energy resolution adequate for the purposes of this disclosure. The (narrow) neutron beam 907 may be scanned across the container 906 sequentially to illuminate discrete columns where the fissile sample 908 may be located. This serves to better localize the position of any fissionable material and will reduce backgrounds from other neutron producing materials in a container. Alternatively, the neutron beam 907 may be a wide fan-like beam encompassing a greater region of the container 906 with the fan opening out in the direction toward the detectors at 90 degrees, for example. This allows a broad scan region of the container but limited in the narrow direction. Such an embodiment would facilitate scanning the container in shorter times for fissile materials. It would detect fissile materials distributed over the dimensions of the fan beam. In this geometry x and y will not be known but they may be inferred from a comparison of the neutron energy spectra on both sides of the container since they should be very close to identical, especially at the highest energies. Starting with any assumption for “a”, such as ½ the width of the container (x=y), the resulting spectra can be adjusted by varying “a” until the spectra are made to have the same high-energy shape. The width of the neutron beam in time T is determined in part by the desired neutron energy resolution. The technology for short duration neutron beam pulses is a well-known art, and pulses of a few nanoseconds are commonly generated by generators of neutron beams. The technology for short duration ion beams used in neutron generators is a well-known art. Time of flight for a 1 MeV neutron over 1 m is 72 nanoseconds. Thus, flight distances of a few meters result in flight times (˜71 nanoseconds for 6 MeV neutrons over a distance of 3 meters, for example) that allow beam pulse duration times of 10 to 20 nanoseconds to separate neutrons of different energies for the purposes of this disclosure. Other specific embodiments are possible and some are mentioned herein as further illustrations of methods to articulate the concepts and methods described earlier. The neutron beams may be of the pulsed variety described above, or they may be of a continuous character as from continuous duty radiofrequency accelerators, DC accelerators or from similarly functioning neutron sources of a monochromatic or nearly monochromatic nature with a well defined highest energy. These embodiments include all such sources to be applied, as anyone skilled in the art will understand. These embodiments include a wide range of neutron beam geometries. One embodiment employs a well-collimated beam scanned across the face of a container. This serves to better localize the position of any fissionable material and will reduce backgrounds from other neutron producing materials in a container. Another embodiment will employ a fan shaped beam allowing a broad scan region of the container but limited in the narrow direction. Such an embodiment would facilitate scanning the container in shorter times for fissile materials. It would detect fissile materials distributed over the dimensions of the fan beam. Another scan embodiment employs a very broad beam geometry in all directions transverse to the beam direction with collimation so as to limit the beam size to that of the container width in its largest manifestation. This embodiment would be very effective in the detection of fissile materials dispersed in small samples over a large volume, such as thin sheets broadly distributed over a large region of the container or small pellets broadly distributed. Many beam geometries are possible, each with specific advantages for certain situations as will be recognized by those skilled in the art, and they are all included in this disclosure. The detection of fissile material can be achieved in one embodiment by measuring the energy or energy distribution of the neutrons from the container. If the energy distribution has neutrons with energies greater than that of the neutron beam used to induce fission, the material in the container has, without question, a fissionable material. The number of neutron signals required to trigger such a determination may be varied according to principles well known in the art to limit rates of false positives while avoiding false negatives. In some embodiments, a single detection event may be considered sufficient; in others a predetermined number may be required; in still others the quantity of neutrons required may be predetermined to vary based on chosen algorithms. The use of angular distributions as well as energy distributions of the neutrons may be included in further determining if the measured energetic neutrons arise from fission. As discussed earlier, the energy distributions and velocity directions of the neutrons are correlated to the directions of the fragments. If the fragment angular distributions are anisotropic, so will the neutron angular distributions be anisotropic. If the fragment angular distributions are isotropic, so will the neutron angular distributions be isotropic. The embodiments of this disclosure use these neutron angular distributions at differing energies of the initiating neutron beam to distinguish between isotopes in the neutron induced fission process. Calibrated anisotropies known to exist at different neutron beam energies for different isotopes will be used. Other embodiments may take advantage of specific features of the neutron-induced fission process for accurate and efficient detection, and to determine the specific actinide present. For example, embodiments using thermal neutrons as incident neutrons may take advantage of filters suitably arranged with respect to the neutron detectors, to absorb thermal neutrons. This increase the likelihood that any neutron that registers in the neutron detector is of higher energy, and thus a fission product. By the same token, the neutron detector can be a threshold detector that does not recognize thermal neutrons. Because prompt neutrons emitted from a fission source in the container may suffer differential absorption along their paths from the container to the neutron detectors, it may be difficult to get a true measure of the angular distribution of the prompt neutrons, such as would enable the determination of the nuclear species involved. In one embodiment, therefore, if fission products are detected using energetic incident neutrons, thermal neutrons are then used as incident neutrons. Since it is known (see above) that the angular distribution of prompt fission neutrons produced from incident thermal neutrons is isotropic, the detection system may be calibrated as to neutron absorption by using the measured angular distribution of the prompt fission neutrons produced by the incident thermal neutrons, and with that calibration information the measured angular distribution of the prompt fission neutrons from the energetic incident neutrons maybe used to determined the nuclear species. Other embodiments may measure the energy distributions of the prompt fission neutrons detected at different angles, to determine the fission species present. Another embodiment takes advantage of the unequivocal fission signal that results from the detection of energetic prompt neutrons from neutron-induced fission, to validate a detection of fissionable material made using a photon beam. Use of a photon beam in the detection process has certain advantages, but there are situations where there may be uncertainty as to whether results are from fission or from a confounding source, such as, for example, 9Be present in the container. To avoid such issues, while maintaining some of the advantages of photon-induced fission for detection, a neutron source can be applied to test for neutron-induced fission in the event a potential fission source is found in a photon-induced fission inspection. Another embodiment removes the uncertainty in the energy distribution and angular dependencies of the prompt neutrons caused by the differential absorption along different paths that neutrons take in traversing a container to the different detectors. This embodiment directs the neutron beam into the container in different directions. For example, in one arrangement the neutron beam enters the container from the top and the neutron detectors view the neutrons at 100 degrees to the beam and at 170 degrees from the beam. By altering the direction of the neutron beam to enter from the side of the container the detectors change roles. The one previously at 100 degrees is now at 170 degrees and the one previously at 170 degrees is now at 100 degrees. However, the differential aspects of neutron absorption remain exactly the same. The two measurements now provide a clear indication of the influence on the neutron energy distribution of the angle of emission of the fission neutrons relative to the neutron beam direction as well as the angular distribution of the neutrons relative to the neutron beam direction. As one particular feature, if the neutron-induced fission process is isotropic the relative neutron yields in the detectors will not change. A change indicates anisotropy of the angular distribution in the original neutron-induced fission process. The change will be a function of the neutron energy and of the nucleus undergoing fission. The embodiments for photon-induced fission with generalized detector angles and photon beam directions and summarized by Equation 1 are also applicable to neutron-induced fission wherein the photon beam is replaced by a neutron beam. All such embodiments are included herein. Neutron detection can be based on reaction energies between the neutrons and the component materials in the detector. Detectors of such a nature may sometimes but not always be called “threshold detectors” because a reaction will occur only if the neutron energy is greater than a certain amount. Examples of such reactions include but are not limited to (n, n′γ), (n, n′f), (n, n′p), (n, n′d) and (n, n′α). Detection of the event may be based on, but not limited to, the detection of: a scintillation event and measuring the deposited energy; the charge created by ionization in a material and measuring the total charge; and, the detection of radioactive nuclei, wherein the radioactivity would be induced only if the neutron energy (energies) were greater than a certain value (or values). All such methods are included in the embodiments described in this disclosure. As discussed above, some commercially available plastic and liquid scintillators can identify neutrons unambiguously using suitable signal processing techniques. Such detectors also have fast enough time response to qualify for the purposes herein and these will be known to those skilled in the art. Such detectors operate in part as proton recoil detectors, based on the energy imparted to protons by the elastic scattering of neutrons from the protons in the hydrogenous material. Therefore, in part, they can function as “threshold detectors” as discussed above, as well as providing the time for an event in a detector and identifying the event as a neutron. Such detection methods are part of the embodiments described herein. Unless otherwise specified, the illustrative embodiments can be understood as providing exemplary features of varying detail of certain embodiments, and therefore, unless otherwise specified, features, components, modules, and/or aspects of the embodiments can be otherwise combined, specified, interchanged, and/or rearranged without departing from the disclosed devices or methods. Additionally, the shapes and sizes of components are also exemplary, and unless otherwise specified, can be altered without affecting the disclosed devices or methods. Other specific embodiments are possible and some are mentioned herein as further illustrations of methods to articulate the concepts and methods described earlier. Although the terms “nuclear material”, “fissionable nuclear material”, “fissile material”, and “fissionable material” have been variously used in this disclosure, the intent of the inventors is that these terms are used interchangeably and are all intended to designate those materials that can be induced to fission by the effect of a gamma ray or by a thermal neutron or fast neutron. These terms are not intended to mean materials that emit neutrons in response to gamma or neutron irradiation, unless such materials also may be induced to fission by the effect of a gamma ray or by a thermal neutron or a fast neutron. The term “container” as used herein is intended to include any enclosure or partial enclosure that may enclose or partially enclose a fissionable material so as to hide or partly hide it or shield it or partly shield it from conventional detection methods—it includes but is not limited to cargo and shipping containers and vehicles. The term “container” also is intended to include any assemblage or other aggregation of material or components with the property that interior portions are not visually observable from the outside, such that x-ray or other techniques (such as those disclosed herein) are required to determine the interior contents and/or composition. The term “determining the energy of a detected neutron”, and similar terms, as used herein is intended to encompass as well determining a lower bound on the energy, or a minimum energy, of the neutron, in contexts where the neutron energy or minimum neutron energy determined is being compared to a predetermined energy in order to establish whether the neutron has energy in excess of a given amount. While the systems and methods disclosed herein have been particularly shown and described with references to exemplary embodiments thereof, it will be understood by those skilled in the art that various changes in form and details may be made therein without departing from the scope of the disclosure. It should be realized this disclosure is also capable of a wide variety of further and other embodiments within the spirit of the disclosure. Those skilled in the art will recognize or be able to ascertain using no more than routine experimentation, many equivalents to the exemplary embodiments described specifically herein. Such equivalents are intended to be encompassed in the scope of the present disclosure.
	description	The present application is a utility application claiming the benefit of provisional application U.S. Pat. Appln. No. 61/835,700, filed on Jun. 17, 2013 by the same inventors, the entirety of which is hereby incorporated by reference. The United States Government has rights in this invention pursuant to Contract No. DE-AC02-06CH11357 between the United States Government and UChicago Argonne, LLC representing Argonne National Laboratory. The present invention relates generally to precision positioning stage systems, and more particularly, relates to a method and high precision robot arm system. A traditional way to perform diffraction measurement is to use a diffractometer. In carrying out diffraction measurement using a small focused beam, or so-called nanodiffraction, using a conventional diffractometer faces three major instrument challenges. First, rotation motion of the sample typically has a large (>10 microns) run-out error known as a sphere of confusion. Second, the circles for the detector are mechanically coupled to the sample, causing unwanted displacement of the sample position when the detector is rotated. Third, some X-ray microscopes used for micro or nanodiffraction use a large vacuum enclosure, making it virtually impossible to use a conventional diffractometer. U.S. Pat. No. 7,331,714 by Deming Shu, Jorg M. Maser, Barry Lai, Stefan Vogt, Martin V. Holt, Curt A. Preissner, Robert P. Winarski, and Gregory B. Stephenson, entitled Optomechanical Structure for a Multifunctional Hard X-Ray Nanoprobe Instrument discloses a multifunctional hard X-ray nanoprobe instrument for characterization of nanoscale materials and devices includes a scanning probe mode with a full field transmission mode. The scanning probe mode provides fluorescence spectroscopy and diffraction contrast imaging. The full field transmission mode allows two-dimensional (2-D) imaging and tomography. The nanoprobe instrument includes zone plate optics for focusing and imaging. The nanoprobe instrument includes a stage group for positioning the zone plate optics. The nanoprobe instrument includes a specimen stage group for positioning the specimen. An enhanced laser Doppler displacement meter (LDDM) system provides two-dimensional differential displacement measurement in a range of nanometer resolution between the zone-plate optics and the sample holder. A digital signal processor (DSP) implements a real-time closed-loop feedback technique for providing differential vibration control between the zone-plate optics and the sample holder. There are many synchrotron radiation instrument applications that require a very high reproducibility for multidimensional linear positioning systems, for example, with nanometer resolution. For X-ray microdiffraction and nanodiffraction studies precision is a key to success. More precise probes and robotic arms are needed to get better results. Thus, it is desirable to provide a method and high precision robot arm system suitable for X-ray nanodiffraction with X-ray nanoprobe. These and other objects, aspects, and advantages of the present disclosure will become better understood with reference to the accompanying description and claims. In brief, a method and high precision robot arm system are provided, for example, for X-ray nanodiffraction with an X-ray nanoprobe. The robot arm includes duo-vertical-stages and a kinematic linkage system. A two-dimensional (2D) vertical plane ultra-precision robot arm supporting an X-ray detector provides positioning and manipulating of the X-ray detector. A vertical support for the 2D vertical plane robot arm includes spaced apart rails respectively engaging a first bearing structure and a second bearing structure carried by the 2D vertical plane robot arm. In accordance with features of the invention, the robot arm system includes a granite base with an air-bearing support, a 2D horizontal base stage, and a vertical axis goniometer with the 2D vertical plane robot arm. In accordance with features of the invention, the robot arm system includes a 3D fast scanning stages group provided with the robot arm. In accordance with features of the invention, the X-ray detector provided with the robot arm includes a 2D X-ray pixel detector. In accordance with features of the invention, the high-precision robot arm system enables effective positioning and manipulating of an x-ray detector for an X-ray nanodiffraction experimental station at a hard X-ray nanoprobe (HXN) beamline. In accordance with features of the invention, the robot arm system includes duo-vertical-stages and a kinematic linkage system providing high load capacity with micron level positioning repeatability. The following description is provided to enable any person skilled in the an to use the invention and sets forth the best mode contemplated by the inventor for carrying out the invention. Various modifications, however, will remain readily apparent to those skilled in the art, since the principles of the present invention are defined herein specifically to provide a method and high precision robot arm system are provided, for example, for X-ray nanodiffraction with an X-ray nanoprobe. For the X-ray microscope for the Hard X-ray nanoprobe (HXN) Beamline at the National Synchrotron Light Source II (NSLS-II) of Brookhaven National Laboratory, which is designed for X-ray microscopy capabilities with an initial spatial resolution of 10 nm, a novel design approach now is taken for nanodiffraction. In the new design, the detector positioning system is completely de-coupled from the sample stage stack together with the use of a high-precision robot arm system in generating full spherical coordinate motions for an array detector. These motion degrees of freedom include the horizontal diffraction angle, the vertical diffraction angle, and the sample-to-detector distance, called detector distance in the following description. When performing Bragg crystallography, adjusting detector distance is important because it defines the maximum range of diffraction angle, which contributes to the special resolution limit of the reconstruction. For example, the instrument requirement for the NSLS-II HXN beamline is to achieve a horizontal diffraction angular range from −5 to 45°, and the sample-to-detector distance from 0.25 to 0.54 m (0.25 to 1.5 m with base station positioning). The horizontal diffraction angle is produced by using a 2-D horizontal base stage plus an in-plane goniometer. The vertical diffraction angle is achieved using a vertical plane robot arm structure of the invention. In addition, a 3-D fast scanning stages group of the invention is used to generate the detector distance plus the capability to raster the area detector, in order to achieve a larger q-range. Having reference now to the drawings, in FIG. 1, there is shown an example high precision robot arm system for X-ray nanodiffraction generally designated by the reference character 100 including an example robot arm generally designated by the reference character 102 together with a robot arm vertical support generally designated by the reference character 104 in accordance with the preferred embodiment. In accordance with features of the invention, the 2D vertical plane robot arm 102 with duo-vertical-stages and a kinematic linkage system provide high load capacity with micron level positioning repeatability. For example, the 2D vertical plane robot arm 102 is 10-20-times better than typical Cartesian robot arm and joints robot arm with similar load capacity. The high precision robot arm system 100 includes the two-dimensional (2D) vertical plane ultra-precision robot arm 102 supporting an X-ray detector 106 to provide positioning and manipulating of the X-ray detector. The X-ray detector 106, for example includes a 2D X-ray pixel detector. While the high precision robot arm system 100 is described particularly for use for X-ray nanodiffraction with an X-ray nanoprobe, it should be understood that the mechanical design may be applied in other fields of precision positioning devices. Other fields of precision positioning devices potentially include additive manufacturing devices (3D printing and sputtering deposition), laser applications, microsurgery, and precision assembly. The vertical support 104 for the 2D vertical plane robot arm 102 includes a pair of spaced apart rails 108, 110 respectively engaging a first bearing structure 202 and a second bearing structure 204 carried by the 2D vertical plane robot arm 102, as shown in FIG. 2. The robot arm system 100 includes a 3D fast scanning stages group generally designated by the reference character 112 provided together with the 2D vertical plane robot arm 102. The 2D vertical plane robot arm 102 carries a robot finger or pin 114 for supporting a sample, for example at an industrial application. In accordance with features of the invention, the high-precision robot arm system enables effective positioning and manipulating of the X-ray detector 106 for an X-ray nanoprobe experimental station at a hard X-ray nanoprobe (HXN) beamline, such as an X-ray nanodiffraction unit 702 as illustrated and described with respect to FIGS. 8 and 9. The robot arm system 100 includes a granite base with an air-bearing support 116, a 2D horizontal base stage 118 including fine positioning adjusters 120, 222. The robot arm system 100 includes a vertical axis goniometer 126 carrying the 2D vertical plane robot arm support 104. The normal load capacity of an example robot arm 102 is 60 kg, which is not only capable to cover the total weight of an array detector 106, such as a Timepix™ QTPX-262K (Amsterdam Scientific Instruments B.V.), and the 3D fast scanning stages group 112 including a pair of motorized stages for the area detector raster scan, but also is capable to carry out the dynamic forces generated during the detector raster scan. The load capacity also provides for positioning loads such as lasers and additive manufacturing devices. Referring also to FIGS. 2, 3, 4, and 5, the detector robot arm 102 is shown. The first bearing structure 202 and second bearing structure 204 carried by the 2D vertical plane robot arm 102 are assembled with a frame member 206 of the robot arm 102. The second bearing structure 204 includes a support member 208 received within a cavity rail or slot 210 formed in the frame member 206. The frame member 206 of the robot arm 102 includes a separate linkage or frame member 212 provided with the 3D fast scanning stages group 112. FIG. 2 shows the robot arm linkage sub-assembly 102. The sub-assembly 102 consists of the frame member 206 providing a linkage base 206, cavity rail or slot 210 forming a high rigidity linear rolling guide 210, and the pair of high rigidity cross-roller bearings 202, 204. The first bearing structure 202 is mounted on the linkage base 206, and the section bearing structure 204 is mounted on the carriage 208 of the rolling guide 210. The linkage sub-assembly 102 is connected to the carriages 108, 110 of FIG. 1, (and 508, 510 of FIG. 6) of the two vertical linear stages through the pair of cross-roller bearings 202, 204. The vertical plane robot arm 102 is positioned in the vertical plane by the two vertical linear stages kinematically. A set of motorized 3-D fast scanning stages group 112 is mounted onto the linkage base to perform the detector raster scan and adjust the detector distance for the 2-D area detector 106. Table 3 summarized the example design specifications of the robot arm vertical linear stages. The linkage or frame member 212 is shown in FIG. 4 is slidingly mounted on a stage base 302. The frame member 212 is shown transparently in FIG. 5 to illustrate the frame member 212 mounted on a carriage 402 after the stage base 302 carrying the frame member 212. In accordance with features of the invention, the detector robot arm 102 is rotatable and slidingly moved along the vertical support member 110 for positioning and manipulating of the X-ray detector 106. Referring also to FIGS. 6, 7, 8, and 9, there is shown another exemplary high precision robot arm system for X-ray nanodiffraction generally designated by the reference character 500 including an example detector robot arm generally designated by the reference character 502 together with a robot arm vertical support generally designated by the reference character 504 in accordance with the preferred embodiment. The high precision robot arm system 500 and the detector robot arm 502 provide similar function and operation as the high precision robot arm system 100 and the detector robot arm 102. In FIG. 6, a side view is shown of the high precision robot arm system 500 including the detector robot arm 502 together with the robot arm vertical support 504 with a top view of the high precision robot arm system 500 shown in FIG. 7. The high precision robot arm system 500 includes the two-dimensional (2D) vertical plane ultra-precision robot arm 502 supporting an X-ray detector 506 to provide positioning and manipulating of the X-ray detector. The robot arm system 500 includes a 3D fast scanning stages group generally designated by the reference character 512 provided together with the 2D vertical plane robot arm 502. The 2D vertical plane robot arm 502 carries a robot finger or pin 514 for supporting a sample, for example at an industrial application. The robot arm system 500 includes a granite base with an air-bearing support 516, a 2D horizontal base stage 518 and a vertical axis goniometer 526 carrying the 2D vertical plane robot arm support 504. In FIGS. 6 and 7, the detector robot arm 502 is illustrated at different positions moved along the vertical support 504 in the operation of the high precision robot arm system 500. FIGS. 8 and 9 schematically illustrates not to scale side and top views an X-ray nanodiffraction apparatus generally designated by the reference character 700 including the exemplary high precision robot arm system 500 together with an X-ray nanodiffraction unit 702 in accordance with a preferred embodiment. In FIGS. 6 and 7, the detector robot arm 502 is illustrated at different positions moved along the vertical support 504 in the operation of the high precision robot arm system 500. The initial design goal for the robot arm unidirectional positioning, for example, robot arm 102, 502 repeatability is 20 micron in the experimental station with +/−0.1 degree temperature control. With a better environmental temperature control, +/−0.1 degree temperature control. With a better environmental temperature control positioning repeatability in a few-micron-level should be achievable for the robot arm system 100, 500. This is valuable for the future detector upgrade with high-spatial-resolution pixel sizes. Table 1 summarizes example main design specifications of the example robot arm systems 100, 500. TABLE 1Example main design specifications for the robot arm systems 100, 500.ParameterValueUnitsOverall dimension1700(W) × 2020(L) × 2750(H)mmHorizontal diffraction−5-+65degangular rangeVertical diffraction angular−5-+45degrangeSample-to-detector distance0.25-0.54mSample-to-detector distance 0.25-2.5 m(with base positioning)Normal load capacity18kgRobot base relocation range2.5 × 3kmUnidirectional positioning<20 micronrepeatability Table 2 summarizes example main design specifications of the 2-D horizontal base stage for the robot arm systems 100, 500. TABLE 2Example design specifications for robot arm 2-D horizontalbase stages 118, 518ParameterValueUnitsOverall dimension1700(W) × 2020(L) × 2750(H)mmNormal Load Capacity500kgStage driver typeStepping motor withplanetary gearheadEncoder typeLinear grating encoderTravel range (mm)1050(W) × 300(L)mmMin.incremental motion (micron)0.25-1.5micronUnidirectional repeatability+/−5 micron(micron)Max. speed (mm/sec)600mm/min Each robot arm system 100, 500 has a unique vertical plane robot arm structure. As shown FIGS. 1 and 2, and in FIG. 6, the vertical plane robot arm structure 104, 504 includes a vertical base 104, 504, two THK™ custom-built vertical linear stages 108, 110; 508, 510 or other commercial stages and a robot arm linkage 102, 502, and a 3-D fast scanning stages group 112, 512 to generate the detector distance plus the capability to raster the area detector 106. TABLE 3Example design specifications for robot arm vertical linear stagesParameterValueUnitsOverall Dimension1520(H) × 200(W) × 58(D)mm(Upstream stage)Overall Dimension1788(H) × 200(W) × 58(D)mm(Downstream stage)Travel range (Upstream stage)850mmTravel range (Downstream1210 mmstage)Vertical Load Capacity 60kgStage driver typeStepping motor withplanetary gearheadEncoder typeLinear grating encoderMin. incremental linear0.8/full-stepmicronmotion0.02/micro-stepMin. incremental angular2.1/full-step micro-motion0.05/micro-stepradUnidirectional repeatability+/−3 micronMax. speed160mm/min In a brief summary, the above describes optomechanical design and example specifications of a high-precision robot arm system 100, 500 to be constructed for the X-ray nano-diffraction experimental station at the HXN beamline for the NSLS-II project. It is expected that its unique vertical plane robot arm structure providing high-precision angular positioning capability for the X-ray nano-diffraction applications with x-ray nanoprobe, will be adopted and used for other X-ray microprobes and nanoprobes to enable high-precision X-ray diffraction capability. It is to be understood that the above-described arrangements are only illustrative of the application of the principles of the present invention and it is not intended to be exhaustive or limit the invention to the precise form disclosed. Numerous modifications and alternative arrangements may be devised by those skilled in the art in light of the above teachings without departing from the spirit and scope of the present invention. It is intended that the scope of the invention be defined by the claims appended hereto. In addition, the previously described versions of the present invention have many advantages, including but not limited to those described above. However, the invention does not require that all advantages and aspects be incorporated into every embodiment of the present invention. All publications and patent documents cited in this application are incorporated by reference in their entirety for all purposes to the same extent as if each individual publication or patent document were so individually denoted.
	claims	1. An inertial confinement fusion reaction chamber comprising:a target insertion mechanism for positioning a directional target, wherein the directional target is configured to produce:a first type of emission primarily in a first direction; anda second type of emission primarily in a second direction;a cylindrical pressure vessel with an outer wall comprising:a first material positioned within the pressure vessel to receive the first type of emission from the directional target in the first direction;a second material positioned within the pressure vessel to receive the second type of emission from the directional target in the second direction; anda third material, the third material comprising a neutron absorption layer disposed within the pressure vessel; anda plurality of beam channels embedded within the outer wall of the pressure vessel. 2. The inertial confinement fusion reaction chamber according to claim 1 wherein the axial length of the cylindrical pressure vessel is 2 times the diameter of the cylindrical pressure vessel. 3. The inertial confinement fusion reaction chamber according to claim 1 wherein the axial length of the cylindrical pressure vessel is 5 times the diameter of the cylindrical pressure vessel. 4. The inertial confinement fusion reaction chamber according to claim 1, wherein the neutron absorption layer has a thickness of 0.2 to 1.5 m. 5. The inertial confinement fusion reaction chamber according to claim 1 further comprising radiation tiles coupled with the inside of the neutron absorption layer. 6. The inertial confinement fusion reaction chamber according to claim 1 further comprising coolant channels disposed within the a neutron absorption layer. 7. The inertial confinement fusion reaction chamber according to claim 1, wherein the neutron absorption layer comprises a neutron moderating material with a neutron absorbing material. 8. The inertial confinement fusion reaction chamber according to claim 1, wherein the beam channels are angled towards the directional target. 9. The inertial confinement fusion reaction chamber according to claim 1, wherein the difference between the outer radius and the inner radius of the cylindrical pressure vessel is 0.2 to 1.5 m. 10. The inertial confinement fusion reaction chamber according to claim 1, wherein some of the plurality of beam channels are disposed cylindrically around the circumference of the cylinder. 11. The inertial confinement fusion reaction chamber according to claim 1 further comprising a tritium breeding mechanism. 12. The inertial confinement fusion reaction chamber according to claim 11, wherein the tritium breeding mechanism comprises channels formed within the pressure vessel that are filled at least in part with lithium. 13. The inertial confinement fusion reaction chamber according to claim 1 further comprising a sacrificial layer disposed on the inner surface of the cylinder. 14. The inertial confinement fusion reaction chamber according to claim 1 further comprising a plurality of injection nozzles configured to deposit a sacrificial layer on the inner surface of the chamber.
	claims	1. A jet pump slip joint clamp comprising:a compression flipper configured to bias against a guide ear of a diffuser in the jet pump;a biasing member configured to bias the compression flipper against the guide ear from an inlet mixer in the jet pump; anda support structure coupled to the inlet mixer and rotatably supporting the compression flipper. 2. The jet pump slip joint clamp of claim 1, wherein the support structure includes,at least one support configured to be rigidly attached to the inlet mixer, anda pivot pin joined to the support, the compression flipper rotatably coupled to the pivot pin. 3. The jet pump slip joint clamp of claim 2, wherein the compression flipper is rotatable about the pivot pin, and wherein the biasing member is configured to bias the compression flipper in a plane of the rotation about the pivot pin. 4. The jet pump slip joint clamp of claim 1, further comprising:a stop pin connected to the support structure, the stop pin positioned about the compression flipper to prevent movement of the compression flipper beyond the stop pin. 5. The jet pump slip joint clamp of claim 1, wherein the biasing member is one of a spring and an elastic rod. 6. The jet pump slip joint clamp of claim 5, wherein the compression flipper includes a divot and wherein the biasing member joins to the compression flipper in the divot. 7. The jet pump slip joint clamp of claim 6, wherein the biasing member is rigidly attached to the compression flipper and not rigidly attached to the inlet mixer. 8. The jet pump slip joint clamp of claim 1, wherein the compression flipper has a curved shape such that the compression flipper is configured to bias against the guide ear at different positions with equal force. 9. A jet pump slip joint clamp, comprising:two supports rigidly affixed to an inlet mixer of the jet pump;a compression flipper rotatably coupled to the two supports; anda biasing member coupled to the compression flipper and biasing the compression flipper away from the inlet mixer of the jet pump. 10. The jet pump slip joint clamp of claim 9, wherein the two supports are affixed to the inlet mixer so as to extend on respective sides of a guide ear of a diffuser of the jet pump and so as to align the compression flipper with the guide ear. 11. The jet pump slip joint clamp of claim 10, wherein the biasing member is not affixed to the inlet mixer and biases the compression flipper against the guide ear, and wherein the compression flipper is joined to the two supports so as to rotate only about an axis perpendicular to the direction in which the two supports extend on the respective sides of the guide ear. 12. The jet pump slip joint clamp of claim 11, wherein the compression flipper has a convex surface in a direction against the guide ear. 13. The jet pump slip joint clamp of claim 12, further comprising:a stop pin connected between the two supports in a path of the rotation of the compression flipper. 14. A stabilizing system for reducing lateral movement or vibration between a diffuser and an inlet mixer of a jet pump, the system comprising:a plurality of slip joint clamps affixed to the inlet mixer and biasing against the diffuser to bias the diffuser of the jet pump apart from the inlet mixer. 15. The system of claim 14, wherein each slip joint clamp includes,at least one support rigidly joined to and extending radially outward from a surface of the inlet mixer at a position of a guide ear of the diffuser,a compression flipper attached to the at least one support and configured to bias against the guide ear, anda biasing member configured to bias the compression flipper against the guide ear from the inlet mixer in the jet pump. 16. The system of claim 15, wherein each slip joint clamp further includes a pivot pin joined to the support, the compression flipper rotatably coupled to the pivot pin. 17. The system of claim 16, wherein each slip joint clamp further includes a stop pin connected to the support in a path of the rotation of the compression flipper about the pivot pin. 18. The system of claim 14, wherein each of the clamps applies a biasing force only in a radial direction to the diffuser. 19. The system of claim 14, wherein each of the slip joint clamps is opposite another slip joint clamp about a central longitudinal axis of the jet pump. 20. The system of claim 14, wherein there are only four slip joint clamps and only four guide ears.
	description	1. Field of the Invention This invention relates to a shipping container for transporting nuclear fuel components and, in particular, to such a container for transporting unirradiated nuclear fuel assemblies and control rod spider assemblies. 2. Related Art In the shipping and storage of unirradiated nuclear fuel elements and assemblies, which contain large quantities and/or enrichments of fissile material U235, it is necessary to assure that criticality is avoided during normal use, as well as under potential accident conditions. For example, nuclear reactor fuel shipping containers are licensed by the Nuclear Regulatory Commission (NRC) to ship specific maximum fuel enrichments; i.e., weights and weight percent U235 for each fuel assembly design. In order for a new shipping container design to receive licensing approval, it must be demonstrated to the satisfaction of the NRC that the new container design will meet the requirements of the NRC's rules and regulations, including those defined in 10 CFR §71. These requirements define the maximum critical accident (MCA) that the shipping container and its internal support structures must endure while maintaining the sub-criticality of the fuel assembly housed therein. U.S. Pat. No. 4,780,268, which is assigned to the assignee of the present invention, discloses a shipping container for transporting two conventional nuclear fuel assemblies having a square top nozzle, a square assembly of fuel rods and a square bottom nozzle. The container includes a support frame having a vertically extending section between the two fuel assemblies which sit side-by-side. Each fuel assembly is connected to the support frame by clamping frames which each have two pressure pads. This entire assembly is connected to the container by a shock mounting frame having a plurality of shock mountings. Sealed within the vertical section are at least two neutron absorber elements. A layer of rubber cork cushioning material separates the support frame and the vertical section from the fuel assemblies. The top nozzle of each of the conventional fuel assemblies is held along the longitudinal axis thereof by jack posts with pressure pads that are tightened down on the square top nozzle at four places. The bottom nozzle of some of these conventional fuel assemblies has a chamfered end. These fuel assemblies are held along the longitudinal axis thereof by a bottom nozzle spacer that holds the chamfered end of the bottom nozzle. These and other shipping containers, e.g., RCC-4, for generally square cross-sectional geometry pressurized water reactor (PWR) fuel assemblies used by the assignee of the present invention, are described in Certificate of Compliance No. 5454, U.S. Nuclear Regulatory Commission, Division of Fuel Cycle and Material Safety, Office of the Nuclear Material Safety and Safeguard, Washington, D.C. 20555. U.S. Pat. No. 5,490,186, assigned to the assignee of the present invention, describes a completely different nuclear fuel shipping container designed for hexagonal fuel, and more particularly, for a fuel assembly design for a Soviet style VVER reactor. Still, other shipping container configurations are required for boiling water reactor fuel. There is a need, therefore, for an improved shipping container for a nuclear fuel assembly that can be employed interchangeably with a number of nuclear reactor fuel assembly designs. There is a further need for such a nuclear fuel assembly shipping container that can accommodate a single assembly in a light-weight, durable and licensable design. These and other needs have been partially resolved by U.S. Pat. No. 6,683,931, issued Jan. 27, 2004 and assigned to the assignee of the instant invention. The shipping container described in this latter patent includes an elongated inner tubular liner having an axial dimension at least as long as the fuel assembly. The liner is preferably split in half along its axial dimension so that it can be separated like a clam shell for placement of the two halves of the liner around the fuel assembly. The exterior circumference of the liner is designed to be closely received within the interior of an overpack formed from an elongated tubular container having an axial dimension at least as long as the liner. Preferably, the walls of the tubular container are constructed from relatively thin shells of stainless steel and the liner is coaxially positioned within the tubular container with close-cell polyurethane disposed in between. Desirably, the inner shell includes boron impregnated stainless steel. The tubular liner enclosing the fuel assembly is slidably mounted within the overpack and the overpack is sealed at each end with end caps. The overpack preferably includes circumferential ribs that extend around the circumference of the tubular container at spaced axial locations that enhance the circumferential rigidity of the overpack and form an attachment point for peripheral shock-absorbing members. An elongated frame preferably of a birdcage design, is sized to receive the overpack within the external frame in spaced relationship with the frame. The frame is formed from axially spaced circumferential straps that are connected to circumferentially spaced, axially oriented support ribs that fixably connect the straps to form the frame design. A plurality of shock absorbers are connected between certain ones of the straps and at least two of the circumferential ribs extending around the overpack, to isolate the tubular container from a substantial amount of any impact energy experienced by the frame, should the frame be impacted. Although the shipping container described in the aforementioned '931 patent is a substantial improvement in that it can accommodate different fuel assembly designs through the use of complementary shell liners while employing the same overpack and birdcage frame, that improvement has been taken one step further by U.S. Pat. No. 6,748,042 assigned to the assignee of the instant invention. The '042 patent describes a transport system that provides a liner and overpack system that will achieve the same objectives as the '931 patent while further improving the protective characteristics of the transport system and the ease of loading and unloading the nuclear fuel components transported therein. The shipping container includes an elongated tubular container, shell or liner designed to support a nuclear fuel product such as the fuel assembly therein. The interior of the tubular liner preferably conforms to the external envelope of the fuel assembly. The exterior of the tubular liner has at least two substantially abutting flat walls which extend axially. In the preferred embodiment, the cross section of the tubular member is rectangular or hexagonal to match the outer envelope of the fuel assembly and three of the corner seams are hinged so that removal of all the kingpins along a seam will enable two of the side walls to swing open and provide access to the interior of the tubular container. The tubular container or liner is designed to seat within an overpack for transport. The overpack is a tubular package having an axial dimension and a cross-section larger than the tubular liner. The overpack is split into a plurality of circumferential sections (for example, two sections, a lower support section and an upper cover, or three sections, a lower support section and two upper cover sections) that are respectively hinged to either circumferential side of the lower support section and joined together when the overpack is closed. The lower support section includes an internal central V-shaped groove that extends substantially over the axial length of the overpack a distance at least equal to the axial length of the tubular liner. Shock mounts extend from both radial walls of the V-shaped groove to an elevation that will support the tubular liner in space relationship to the groove. The axial location, number, size and type of shock mount employed are changeable to accommodate different loadings. The tubular liner is seated on the shock mounts, preferably with a corner of the liner aligned above the bottom of the V-shaped groove. The top cover section (sections) of the overpack has a complementary inverted V-shaped channel that is sized to accommodate the remainder of the tubular liner with some nominal clearance approximately equal to the spacing between the lower corner of the tubular liner and the bottom of the V-shaped groove. The ends of the overpack are capped and the overpack sections are latched. Though the transport system of the '042 patent provides a substantial improvement in the protective characteristics and ease of loading and unloading of the nuclear components being transported, further improvement in the ease of loading and unloading the liner is desired. The invention of U.S. Pat. No. 7,474,726 assigned to the assignee of the current invention, was conceived for such purpose. The '726 patent is a variation on the '042 patent that permits loading of the liner from the top as well as the sides. The liner comprises an elongated tubular container that has at least two substantially flat walls with at least one circumferential end having a hinged interface with a stationary wall of the container to provide access through the side of the container. The hinged wall extends axially in a direction of one end of the container and terminates a preselected distance short of the corresponding end of the stationary wall. The stationary wall has a lateral groove on an interior surface thereof at an elevation starting substantially at the elevation of the one end of the hinged wall. An access cover is slidable in the groove in the stationary wall to close off the one end of the container so that the interior of the container may be accessed either through the one end by sliding out the access cover, or from the side by rotating the hinged wall. The access cover can be locked in position and the elongated tubular container has the other end opposite the one end capped and sealed and the tubular container is sized to fit within the overpack of the '042 patent. While the '726 patent is a further improvement in the design of shipping casks which provides greater versatility in the number of fuel assembly designs that can be accommodated and the ease of loading, further improvement is still desired to accommodate an increased number of fuel assembly and component designs. Accordingly, it is an object of this invention to further enhance the versatility of the shipping cask design described in the '726 patent. It is an additional object of this invention to enhance the shipping cask design of the '726 patent so that it will accommodate all Westinghouse designed light water reactor nuclear fuel assemblies. It is an additional object of this invention to further improve the shipping cask design of the '726 patent so that it will accommodate all CE designed light water reactor nuclear steam supply system fuel assemblies. Furthermore, it is an object of this invention to further enhance the design of the '726 patent so that it will satisfy European and U.S. fuel transport regulations. Furthermore, it is an object of this invention to further enhance the unirradiated nuclear fuel transport cask design of the '726 patent so that it will transport control rod spider assemblies in addition to nuclear fuel assemblies. Additionally, it is an object of this invention to enhance the design of the '726 patent to provide a larger access envelope for the loading of a nuclear fuel assembly into the liner. These and other objects are achieved by a shipping container system for transporting an unirradiated nuclear fuel product, that has an elongated tubular container with an access opening that extends over an elongated dimension of the tubular container and is designed to receive and support the nuclear fuel product. An interior of the tubular container has at least two substantially flat, movable walls and at least two substantially flat, stationary walls with a circumferential end of the at least two stationary walls connected together along the elongated dimension and another circumferential end of at least one of the stationary walls having a hinged interface with one circumferential end of at least one of the at least two movable walls. Another circumferential end of the at least one of the at least two movable walls is connected to one circumferential end of another of the at least two movable walls with another circumferential end of the another of the at least two movable walls connectable to another circumferential end of another of the at least two stationary walls. Each of the stationary walls and the movable walls have upper and lower ends and the stationary walls have at least one of either a bar or groove on an interior side of at least two of the stationary walls, proximate the upper end, that extend along at least a portion of the corresponding stationary walls substantially in a direction orthogonal to the axis of the elongated tubular container. The elongated tubular container also has a top plate that closes off a top of the container, with the top plate having at least two peripheral sides having the other of the one of a bar or groove extending substantially along at least a portion of an outer edge with the other of the one a bar or groove sized and oriented to mate with the corresponding one of the bars or grooves. The top plate also has an anchoring mechanism for supporting a side of the top plate against an abutting side of the one or the another of the at least two stationary walls. The shipping container system also includes an elongated, tubular overpack having an axial dimension at least as long as the tubular container and an internal cross section larger than the tubular container. The overpack has an interior tubular channel having an axially extending lower support section supporting a plurality of shock mounts, with at least one of the plurality of shock mounts positioned on either radial side of the support section. The shock mounts support at least one of the flat walls of the tubular container in spaced relationship with the lower support section when the overpack is supported in a horizontal position. At least one circumferential end of the lower support section has a clamp interface substantially along the axial dimension thereof, to provide access to the interior of the overpack. Additionally, means are provided for supporting the overpack in a horizontal position. In one embodiment, the anchoring mechanism is a placement rod having one end with a male locking contour which extends through a wall in the top plate and into an abutting opening in the one or the another of the at least two stationary walls, that has a complimentary female locking contour. Preferably, the male locking contour is a male threaded end of the placement rod and the complimentary female locking contour is a female thread on an interior surface of the abutting opening. Desirably, a hole in a wall of the top plate through which the placement rod extends includes a female thread that mates with the male thread on the placement rod and the male threaded end of the placement rod is tapered. In still another embodiment, the one of the bars or grooves extends substantially across a full width of each of the walls and the another of the bars or grooves on the edge of the top plate extends substantially around the entire edge. Preferably, the top plate includes at least two spaced openings extending through the top plate and through which push rods extend from above the top plate to an upper surface of the nuclear fuel product being transported. An axial length of the push rods within the interior of the elongated tubular container is adjustable from above the top plate on an exterior of the elongated tubular container. Desirably, the push rod adjustment includes a nut supported above the top plate on each push rod that is retained on the corresponding push rod by a locking pin. Preferably, the spaced openings through the top plate have female threads that mate with male threads on the push rods. Desirably, the top plate also includes a third spaced opening with the first two spaced openings located proximate diametrically opposed corners of the top plate and the third spaced opening located substantially in the center of the top plate. The top plate also includes motion sensors supported on its upper surface for recording the extent of excessive motion of the elongated tubular container. The sensors can be read from outside of the container without opening any of the walls. In addition, the elongated tubular container is preferably sized to transport both a nuclear fuel assembly and a control rod spider assembly. The overpack and internal components of the nuclear fuel product containment and transport system of this invention is an improvement on such systems described in U.S. Pat. Nos. 6,748,042 and 7,474,726, generally illustrated in FIG. 1. The system 10 includes a tubular container or shell 26 (sometimes referred to as a liner) constructed from a material such as aluminum that houses a nuclear fuel assembly and is suspended over a V-shaped groove 32 in an overpack 12, supported on shock mounts 28 that are affixed in a recess 30 in an upper wall section of the groove 32 and spaced along the axial length of the lower overpack section 16. The shock mounts can be those identified by part number J-34324-21, which can be purchased from Lord Corporation having offices in Cambridge Springs, Pa. Angle irons 24 can be used at the corners of the tubular container 26 to spread the load on the container walls. The number and resiliency of the shock mounts are chosen to match the weight of the container, which depends upon the nuclear product being transported within the container 26. The orientation of the lower section 16 of the overpack 12 is fixed by the legs 18 so that the weight of the container 26 holds the container centered in the groove 32. One capped end 22 forms part of the lower overpack support section 16, while a second capped end 20 is formed as an integral part of the top cover 14. The end 20 of the upper overpack segment 14 seals against the lip 21 in the lower support section 16. Similarly, though not shown, the end 22 formed as an integral part of the lower support section 16 seals against a corresponding lip on the upper overpack section 14 in the same manner. Keys 50 are on each side of the upper section 14 of the overpack 12 fit in complimentary key ways in the lower overpack support section 16. The liner or shell 26 illustrated in FIG. 1 is shown in more detail in FIG. 2. The liner 26 includes two stationary walls 34 and 36 that are rigidly coupled along a longitudinal seam 38 at one corner, such as by welding, though it should be appreciated that the walls 34 and 36 can be extruded as a single piece or rigidly connected in some other manner. Two movable walls 40 and 42 are respectively, hingedly connected to the stationary walls 36 and 34 at the hinged seams 54 and 56. The moveable wall 40 is formed in upper and lower halves 46 and 44 and the other movable wall 42 is similarly formed in upper and lower halves 52 and 48. The movable doors 40 and 42 may be hingedly connected as shown by the hinged seam 58 in the lower door 44. Alternatively, the movable walls 40 and 42 may be rigidly connected as shown by the fixed seam 60 in the upper doors 46 and 52. Alternatively, the movable walls may merely be latched together as an indicated by the latch plate 62. A square fuel assembly 64 is shown supported within the interior of the liner 26. Push rods 66 and 68 exert pressure on diagonally opposite corners of the top fuel assembly nozzle 70 to secure the fuel assembly axially as well as radially. Alignment pins 72 and 74 are part of the fuel assembly 64 and serve to align the fuel assemblies with an upper core plate in a nuclear reactor. The push rods 66 and 68 include adjustment nuts 96 that are employed to adjust the length of the push rods 66 and 68 to securely pressure the fuel assembly 64 after it is loaded within the liner 26. Accelerometers 76 are supported on the inside of the upper wall 46 to record any shocks the liner 26 may receive. As noted in the '042 patent, additional stationary walls may be added to accommodate a hexagonal fuel assembly or a fuel assembly having another geometry. The liner illustrated in FIG. 2 is capped at its upper end by a top plate 78 that has a peripheral lip 80 surrounding its outer edge that mates with a corresponding groove 82 in the upper portion of the liner 26. The lip 84 above the groove 82 has proved to be an obstruction to loading certain fuel assembly designs. The spring 86 that is cantilevered over the edge of the top plate 78 is provided to retain the top plate with the movable upper wall portions 46 and 52 when the upper wall portions are swung to the open position. FIG. 3 illustrates the improvement that this invention contributes to the design shown in FIG. 2. Like reference characters are employed to identify the corresponding components. The improvement of this invention provides a newly designed top shear plate 78 that replaces the top plate 78 and wall interfaces 82 and 84 previously described with respect to FIG. 2. Referring to FIG. 3 it will be appreciated that the positioning of the hold down or push rods 66 and 68 and the location of the accelerometers 76 have changed to facilitate ease of use and make the liner 26 compatible for all Westinghouse pressurized water reactor nuclear steam supply system and non-Westinghouse pressurized water reactor nuclear steam supply system fuel assembly designs, including CE, ATOM and B&W fresh fuel assembly designs. The improvement of this invention further comprises a machined aluminum top sheer plate designed to fit the liner previously described, which maintains the fuel assembly axial, lateral and vertical position during transport. The shear top plate has three universal threaded holes to permit transport of Westinghouse and non-Westinghouse pressurized water reactor nuclear fuel assembly designs mentioned previously. FIG. 4 provides a better view of the shear top plate 78 of this invention affixed at the top of the two stationary walls 34 and 36 of the liner 26. The upper portions 46 and 52 of the movable walls 40 and 42 have been removed for clarity. The fuel assembly 64 is shown within the liner with the control rods of a control rod spider assembly inserted within the guide thimbles of the fuel assembly 64. The shear plate 78 has a bordering rail 102 that extends along the periphery of the shear top plate 78. The bordering rail 102 has a groove 90 that extends around its peripheral edge and mates with a horizontal bar or rail 88 on each of the liner walls 34, 36, 40 and 42. The bars 88 are affixed to the liner walls near the top of the liner and may extend across all or part of the width of each wall and is sized to mate with the groove 90 in the peripheral edge of the shear top plate 78. Push bars (also referred to as hold down bars) 66 and 68 extend from the corners of the top nozzle 70 of the fuel assembly 64, upward and through diametrically opposed corner openings 106 and 108 in the top shear plate 78 and are locked in position by adjustment nuts 96. The openings 106 and 108 are threaded and mate with corresponding male threads on the push rods 66 and 68 near the upper end thereof. Cotter pins 98 prevent the adjustment nuts 96 from unscrewing from the ends of the push rods 66 and 68. The push rods 66 and 68 can be tightened down against the diametrically opposite corners of the fuel assembly top nozzle 70 by applying a torque wrench to the ends of the push rods above the shear plate 78 and then tightening down the nuts 96. Accelerometers 76 are affixed to the upper surface of the top shear plate 78 and can be read without opening the liner 26. A placement rod 92 extends through an opening in the border rail 102 and into a hole in the stationary wall 36. The engagement end 100 of the placement rod 92 has a male thread which mates with a female thread in the opening in the border rail 102 and the aligned hole in the wall 36 (not shown). Preferably, the thread on the end 100 of the placement rod 92 is tapered. A third hole 104 is provided through the center of the top shear plate 78 to accommodate a third push rod 110 that can be used to further secure the fuel assembly 64 or other fuel components such as the control rod spider assembly 94. FIGS. 5 and 6 provide isometric views of the top shear plate and its components, taken from different angles. The views shown in FIGS. 5 and 6 provide a better understanding of the components of the top shear plate 78 of this invention. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular embodiments disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the appended claims and any and all equivalents thereof.
	summary
054065976	summary	TECHNICAL FIELD The present invention relates generally to nuclear reactors, and, more specifically, to an improved boiling water reactor. BACKGROUND ART A conventional boiling water reactor (BWR) includes a pressure vessel containing a nuclear reactor core above which are disposed in turn conventional steam separators and dryers. The vessel is filled with a cooling and moderating fluid such as water to a predetermined normal water level located generally near the middle of the steam separators. The core boils the water for generating a steam-water mixture which rises upwardly into the steam separators, which remove some of the water therefrom, with additional water being further removed therefrom from the steam dryers positioned above the steam separators. The dried steam is conventionally discharged from the vessel to a conventional steam turbine, for example, which powers an electrical generator for generating electrical power provided to an electrical utility grid. A typical BWR is controlled by a plurality of control rods which extend downwardly from the core through conventional guide tubes extending from the bottom of the core to the lower head of the pressure vessel which defines therebetween a lower plenum. Extending downwardly from the lower head are a plurality of conventional control rod drives (CRDs) which are effective for selectively inserting the control rods upwardly into the core for reducing reactivity therein, and for selectively withdrawing the control rod downwardly from the core for increasing reactivity therein. Accurate intermediate positions of the control rods may be obtained by using a conventional drive screw which is selectively rotated in opposite directions by a conventional stepper motor to selectively translate upwardly and downwardly a ball nut threadingly engaged therewith. An elongate piston rests on the ball nut and is coupled to a respective control rod for raising and lowering the control rod as the ball nut is correspondingly translated. In order to obtain relatively instantaneous insertion of the control rods during a SCRAM operation, a pressurized fluid such as water is conventionally channeled through the CRD for lifting the piston and in turn lifting the control rod independently of the ball nut. In order to fully withdraw the control rods from below the core, the guide tubes extending between the core and the vessel lower head must have a vertical height approximately equal to the length of the control rods. The height of the core also has a vertical height approximately equal to the length of the control rods so that the control rods may be fully inserted therein. The conventional steam separators additionally require a suitable vertical height for effectively separating water from the steam-water mixture. And additional vertical height is required for the steam dryer disposed above the steam separators. Accordingly, the overall height of the pressure vessel must be suitable for containing these several components and for allowing the effective functioning thereof. A typical pressure vessel for a BWR sized for generating steam to power a turbine-generator for providing electrical power to the electrical utility grid is about 21 meters tall, with the reactor generating on the order of about 1,000 megawatts electric (MWe) and higher. Such a large pressure vessel, which is typically made from steel, has a correspondingly high weight requiring large cranes for the assembly thereof into a power plant. A conventional BWR typically includes conventional recirculation pumps which operate for channeling downwardly the water within the pressure vessel in a conventional annular downcomer surrounding the core, which recirculated water enters the lower plenum and flows upwardly through the core. Since the water used to generate the reactor steam also cools the reactor, systems are typically provided to ensure that adequate water is always contained within the pressure vessel and above the core during all modes of operation of the reactor, including abnormal modes such as that occurring in a conventional loss-of coolant accident (LOCA) wherein the coolant water leaks from the reactor system and must be suitably replaced for maintaining an adequate level of water within the pressure vessel above the core. In one type of advanced BWR, a gravity-driven cooling system (GDCS) includes a pool of water located outside the pressure vessel at an elevation above the reactor core to provide makeup water in a LOCA situation for example. In order to use the GDCS makeup water, the reactor pressure vessel must be first depressurized in a conventional manner to sufficiently reduce the pressure therein so that the pressure head of the elevated GDCS makeup water is sufficient to force the makeup water into the vessel to supplant the lost reactor water for maintaining the reactor water level above the core. Since depressurization of the pressure vessel takes several minutes, the vessel continues to lose its coolant water either as a liquid or from the steam being generated and discharged therefrom, which loss of water must be suitably made up to ensure an adequate water level within the vessel. One arrangement for ensuring adequate water level within the vessel is to provide a greater initial amount of water in the pressure vessel above the core by suitably increasing the normal elevation of the water level within the vessel. By initially providing more water within the vessel, adequate reserves of the water therein may be maintained during a LOCA situation until the vessel may be suitably depressurized and makeup water provided thereto from the GDCS pool. The increased normal water level within the vessel, however, requires a corresponding increase in the height of the pressure vessel, which correspondingly increases its manufacturing complexity and weight. Furthermore, in another abnormal situation involving an accidental trip of all the recirculation pumps, recirculation of the coolant water within the vessel will occur solely by natural recirculation flow of the water therein with the core-heated water rising, and the relatively cooler water within the downcomer falling. By increasing the normal water level as described above. The natural recirculation flow of the coolant water within the vessel is also increased, which is effective for providing additional margin against conventionally known nuclear-thermal-hydraulic instability of the coolant water following an all-pump trip. Furthermore, the increased normal water level is also effective for improving conventional thermal margins and peak pressures for other types of plant operating transient conditions. Analysis indicates that an increase in the normal water level within the pressure vessel of about 7 meters is required both to apply an effective gravity-driven cooling system in a LOCA situation, and to achieve suitably stable operation following an all recirculation pump trip situation for a reactor sized for generating about 1350 MWe. However, in order to provide the additional 7 meters of water above the reactor core, the entire pressure vessel must be extended 7 meters above the core which would increase the normal length thereof from about 21 meters to at least 28 meters. Such a large pressure vessel is near the current fabrication limits, and near the current crane capacity limits for assembling the vessel in the power plant. The relatively large pressure vessel increases the complexity and cost of its use within the power plant. OBJECTS OF THE INVENTION Accordingly, one object of the present invention is to provide a new and improved boiling water reactor (BWR). Another object of the present invention is to provide an improved BWR having an increased normal water level above a reactor core therein without correspondingly increasing the length of the pressure vessel. Another object of the present invention is to provide an improved BWR having an increased normal water level therein with bottom-mounted control rod drives having gravity-aided SCRAM capability. Another object of the present invention is to provide an improved BWR having an increased normal water level above the core thereof which concurrently provides space for withdrawal of the control rods from the core and for providing guidance thereof. DISCLOSURE OF INVENTION A boiling water reactor includes a pressure vessel containing a reactor core, chimney, steam separator assembly, and steam dryer assembly therein, with the vessel being filled with reactor water to a normal water level through the steam separator assembly. A plurality of control rod drives extend downwardly from the bottom of the pressure vessel and are operatively joined to control rods extending upwardly into the reactor core. The chimney includes a plurality of channels disposed above the core and laterally spaced apart to define guide slots for receiving the control rods as they are selectively translated upwardly out of the core by the control rod drives. The chimney has a vertical height for increasing the normal water level above the reactor core and for providing a space for the control rods withdrawn from the reactor core by the bottom-mounted control rod drives. In a preferred embodiment, the control rods are selectively withdrawn upwardly from the core and inserted downwardly into the core by the control rod drives, which also are effective for selectively releasing the control rods for allowing gravity to insert the control rods into the core.
	abstract	Configurations of molten fuel salt reactors are described that allow for active cooling of the containment vessel of the reactor by the primary coolant. Furthermore, naturally circulating reactor configurations are described in which the reactor cores are substantially frustum-shaped so that the thermal center of the reactor core is below the outlet of the primary heat exchangers. Heat exchanger configurations are described in which welded components are distanced from the reactor core to reduce the damage caused by neutron flux from the reactor. Radial loop reactor configurations are also described.
	abstract	A hydraulic operated system for testing cable under variable tensions and speeds with a lower input horsepower requirement. The system can include storage reels in series with multiple moveable sheaves. A moveable tensioning moveable sheave can be in series with the multiple moveable sheaves for receiving cable and for measuring load and speed of the cable using sensors. The system can include clutches and a clutch controller. The measured load and speed can be transmitted to a processor which can perform calculations and comparisons thereon, and can transmit related notifications to client devices through a network.
045284547	summary	CROSS REFERENCE TO RELATED APPLICATIONS The present application is related to commonly assigned copending application Ser. No. 120,108 filed Feb. 8, 1980 (now U.S. Pat. No. 4,274,007 of June 16, 1981 and Ser. No. 966,951 filed Dec. 6, 1978 (now U.S. Pat. No. 4,278,892 of July 14, 1981 and making reference to then-pending applications Ser. No. 940,856 of Sept. 8, 1978, (now U.S. Pat. No. 4,272,683 of June 9, 1981, Ser. No. 940,098 (now U.S. Pat. No. 4,234,798), and Ser. No. 107,276 of Sept. 26, 1979 (now U.S. Pat. No. 4,288,698 of Sept. 8, 1981. Reference may also be had to U.S. Pat. Nos. 4,229,316 and 4,235,739 issued on still earlier applications commonly owned herewith. For the construction of the vessel and as to radiation-shielding properties thereof and the use of such vessels, these prior art applications and patents are hereby incorporated by reference in their entirety and it is noted that the prior art known to applicants to be the most relevant is the art of record in said applications. FIELD OF THE INVENTION As is pointed out in the aforementioned copending applications, it is known to provide for the transport and storage of radioactive wastes, containers or vessels of a radiation-shielding material and which may be provided with channels or compartments to contain radiation-blocking or radiation-attenuating materials, and with ribs or the like to promote heat exchange with ambient air. Radioactive material can be placed in such containers and sealed by cover arrangements of which the most pertinent is that found in application Ser. No. 120,128, in which it is pointed out that an effective closure for the vessel can be provided by forming the mouth of the vessel with a seat receiving the plug-type inner cover having a frustoconical portion and a cylindrical portion fitting into correspondingly shaped parts of the seat and sealed relative to the latter with elastomeric seals, generally O-rings. Above this inner cover an outer cover was provided which extended beyond the outline of the inner cover and was secured to the vessel. The container vessel, which is formed at its upper and lower ends with thickened portions for reinforcement and stability, generally in the form of annular beads or enlargements, can be composed of cast iron and especially spherulitic (nodular) cast iron, can be used for the storage and disposal of radioactive materials of various types, especially irradiated nuclear fuel elements upon their removal from the core of a nuclear reactor. As can be seen from German patent document No. 28 37 631 (see also U.S. Pat. No. 4,278,892), it is frequently advantageous to provide at least one bore in the wall of the vessel, extending from the upper end thereof to open into the interior of the vessel close to the bottom, which serves to feed material into or draw material from the interior of the vessel and/or for control or monitoring purposes. In earlier arrangements utilizing such a bore, the latter terminated at the upper end of the vessel outwardly of the outline of the radiation-shielding or plug-type cover and required special closures. This, in turn, made sealing difficult and complicated the problem of controlling the storage of the radiation material by making access to the material through the bore considerably more difficult. OBJECTS OF THE INVENTION It is the principal object of the present invention to provide an improved storage and transport container for radioactive materials which is free from the disadvantages described. Another object of the invention is to provide an improved container which extends the principles of our U.S. Pat. Nos. 4,274,007 and 4,278,892 mentioned above. It is also an object of the invention to provide a radiation-shielding transport and/or storage container which facilitates control and monitoring of the state of the contents of the vessel. A further object of the invention is the provision of an improved cover structure for a vessel to be utilized for the aforedescribed purposes. SUMMARY OF THE INVENTION These objects and others are attained, in accordance with the present invention, in a transport and/or storage container adapted to receive radioactive materials and in which the wall bore of the vessel terminates at the upper end and within the outline of the shielding or plug-type cover and preferably within the outline of the plug portion thereof, while this cover is formed with a bore adapted to be aligned with the wall bore and in which an obturating element (e.g. a valve or plug) can be fitted to extend into the wall bore and seal the latter, the bore of the cover being covered in turn by an additional or safety cover. In other words, when a plurality of wall bores are provided in the vessel of cast iron, each of them open in the region of the seat into which the plug-forming radiation-shielding cover fits to be aligned with corresponding bores in this cover. In this manner, the shielding cover can be formed in the region of the seat or socket with control, monitoring and test bores in which control or monitoring devices, fittings, valves or the like can be inserted so that these devices, if left in place, or the control or test bores can be covered by the outer of second cover. The number and type of control and test bores will depend upon the test required during transport and storage of the container as well as upon special purposes to which the container may be put and may be determined by a control or monitoring program, by regulation or by statute. Preferably, the additional cover does not sit directly upon the edge of the vessel but rather is recessed therein, i.e. fits into a recess forming a seat for this cover and provided in the mouth-forming end of the vessel so that the outer cover does not project beyond the vessel wall but is either recessed inwardly from this end of the vessel or is flush therewith. Naturally the vessel can be formed with a shoulder or seat against which the additional cover is sealingly set. It is possible to provide still further covers as described, for example, in our concurrently filed copending application Ser. No. 243,562 (now U.S. Pat. No. 4,450,042), for greater security or to facilitate monitoring the seals of the covers by monitoring a control gas which can be introduced into the vessel with the radioactive material as described in the latter application. The reinforcing bead or thickened end of the mouth of the vessel permits the system of the invention to be accommodated economically and without loss of radiation-shielding effect.
	description	This application is a division of U.S. patent application Ser. No. 15/797,627 filed on Oct. 30, 2017, now U.S. Pat. No. 10,629,313, which is a division of U.S. patent application Ser. No. 13/405,405 filed on Feb. 27, 2012, now U.S. Pat. No. 9,805,832, the disclosures of which are hereby incorporated by reference in their entirety. The following relates to the nuclear reactor arts, nuclear power generation arts, nuclear reactor control arts, nuclear reactor electrical power distribution arts, and related arts. In nuclear reactor designs of the integral pressurized water reactor (integral PWR) type, a nuclear reactor core is immersed in primary coolant water at or near the bottom of a pressure vessel. In a typical design, the primary coolant is maintained in a subcooled liquid phase in a cylindrical pressure vessel that is mounted generally upright (that is, with its cylinder axis oriented vertically). A hollow cylindrical central riser is disposed concentrically inside the pressure vessel. Primary coolant flows upward through the reactor core where it is heated, rises through the central riser, discharges from the top of the central riser, and reverses direction to flow downward back toward the reactor core through a downcomer annulus. The nuclear reactor core is built up from multiple fuel assemblies. Each fuel assembly includes a number of fuel rods. Control rods comprising neutron absorbing material are inserted into and lifted out of the reactor core to control core reactivity. The control rods are supported and guided through control rod guide tubes which are in turn supported by guide tube frames. In the integral PWR design, at least one steam generator is located inside the pressure vessel, typically in the downcomer annulus, and the pressurizer is located at the top of the pressure vessel, with a steam space at the top most point of the pressure vessel. Alternatively an external pressurizer can be used to control reactor pressure. A set of control rods is arranged as a control rod assembly that includes the control rods connected at their upper ends with a spider, and a connecting rod extending upward from the spider. The control rod assembly is raised or lowered to move the control rods out of or into the reactor core using a control rod drive mechanism (CRDM). In a typical CRDM configuration, an electrically driven motor selectively rotates a roller nut assembly or other threaded element that engages a lead screw that in turn connects with the connecting rod of the control rod assembly. In some assemblies, such as those described in U.S. Pat. No. 4,597,934, a magnetic jack may be used to control movement of one or more control rods. Control rods are typically also configured to “SCRAM”, by which it is meant that the control rods can be quickly released in an emergency so as to fall into the reactor core under force of gravity and quickly terminate the power-generating nuclear chain reaction. Toward this end the roller nut assembly may be configured to be separable so as to release the control rod assembly and lead screw which then fall toward the core as a translating unit. In another configuration, the connection of the lead screw with the connecting rod is latched and SCRAM is performed by releasing the latch so that the control rod assembly falls toward the core while the lead screw remains engaged with the roller nut. See Stambaugh et al., “Control Rod Drive Mechanism for Nuclear Reactor”, U.S. Pub. No. 2010/0316177 A1 published Dec. 16, 2010 which is incorporated herein by reference in its entirety; and Stambaugh et al., “Control Rod Drive Mechanism for Nuclear Reactor”, Intl Pub. WO 2010/144563 A1 published Dec. 16, 2010 which is incorporated herein by reference in its entirety. The CRDMs are complex precision devices which require electrical power to drive the motor, and may also require hydraulic, pneumatic, or another source of power to overcome the passive SCRAM release mechanism (e.g., to hold the separable roller nut in the engaged position, or to maintain latching of the connecting rod latch) unless this is also electrically driven. In existing commercial nuclear power reactors, the CRDMs are located externally, i.e. outside of the pressure vessel, typically above the vessel in PWR designs, or below the reactor in boiling water reactor (BWR) designs. An external CRDM has the advantage of accessibility for maintenance and can be powered through external electrical and hydraulic connectors. However, the requisite mechanical penetrations into the pressure vessel present safety concerns. Additionally, in compact integral PWR designs, especially those employing an internal pressurizer, it may be difficult to configure the reactor design to allow for overhead external placement of the CRDMs. Accordingly, internal CRDM designs have been developed. See U.S. Pub. No. 2010/0316177 A1 and Intl Pub. WO 2010/144563 A1 which are both incorporated herein by reference in their entireties. However, placing the CRDMs internally to the reactor vessel requires structural support and complicates delivery of electrical and hydraulic power. Electrical conductors that are usable inside the pressure vessel are generally not flexible and are not readily engaged or disengaged, making installation and servicing of internal CRDM units challenging. Disclosed herein are improvements that provide various benefits that will become apparent to the skilled artisan upon reading the following. In one illustrative embodiment, an apparatus is disclosed comprising a plurality of control rod drive mechanisms (CRDMs) each configured to raise or lower a control rod assembly and a distribution plate configured to be mounted in a nuclear reactor pressure vessel and including a plurality of connection sites at which the CRDMS are mounted, the distribution plate including electrical power distribution lines disposed on or in the distribution plate for distributing electrical power to the CRDMs mounted on the distribution plate. A method is also disclosed comprising installing a CRDM in a nuclear reactor by operations which include attaching the CRDM to a top plate of a standoff and connecting a mineral insulated cable between the CRDM and an electrical connector disposed in or on a bottom plate of the standoff to form a CRDM/standoff assembly and mounting the bottom plate of the CRDM/standoff assembly to a distribution plate wherein the mounting connects an electrical power line disposed on or in the distribution plate with the electrical connector disposed in or on the bottom plate of the standoff. In another illustrative embodiment, an apparatus is disclosed comprising a nuclear reactor including a core comprising a fissile material disposed in a pressure vessel, a mechanical reactor component disposed inside the pressure vessel and having a mounting flange with a power connector, and a power distribution plate disposed inside the pressure vessel and having a connection site configured to mate with the flange of the mechanical reactor component, the connection site including a power connector configured to mate with the power connector of the flange of the mechanical reactor component when the flange of the mechanical reactor component is mated with the connection site, power lines on or in the power distribution plate being arranged to deliver power to the power connector of the connection site, wherein the flange of the mechanical reactor component is mated with the connection site of the power distribution plate. FIG. 1 illustrates an integral Pressurized Water Reactor (integral PWR) generally designated by the numeral 10. A reactor vessel 11 is generally cylindrical and contains a reactor core 1, a steam generator 2, and a pressurizer 3. Although a pressurized water reactor (PWR) is depicted, a boiling water reactor (BWR) or other type of nuclear reactor is also contemplated. Moreover, while the disclosed rapid installation and servicing techniques are described with reference to illustrative internal CRDM units, these techniques are readily adapted for use with other internal nuclear reactor components such as internal reactor coolant pumps. In the illustrative PWR, above the core 1 are the reactor upper internals 12 of integral PWR 10, shown in inset. The upper internals 12 are supported by a mid flange 14, which in the illustrative embodiment also supports internal canned reactor coolant pumps (RCPs) 16. More generally, the RCPs may be external pumps or have other configurations, and the upper internals may be supported otherwise than by the illustrative mid flange 14. The upper internals include control rod guide frames 18 to house and guide the control rod assemblies for controlling the reactor. Control Rod Drive Mechanisms (CRDMs) 20 raise and lower the control rods to control the reactor. In accordance with one embodiment, a CRDM distribution plate 22 supports the CRDMs and provides power and hydraulics to the CRDMs. A riser transition 24 directs coolant flow upward. Control rods are withdrawn from the core by CRDMs to provide enough positive reactivity to achieve criticality. The control rod guide tubes provide space for the rods and interconnecting spider to be raised upward away from the reactor core. The CRDMs 20 require electric power for the motors which move the rods, as well as for auxiliary electrical components such as rod position indicators and rod bottom sensors. In some designs, the force to latch the connecting rod to the lead screw, or to maintain engagement of the separable roller nut, is hydraulic, necessitating a hydraulic connection to the CRDM. To ensure passive safety, a positive force is usually required to prevent SCRAM, such that removal of the positive force initiates a SCRAM. The illustrative CRDM 20 is an internal CRDM, that is, is located inside the reactor vessel, and so the connection between the CRDM 20 and the distribution plate 22 is difficult to access. Servicing of a CRDM during a plant shutdown should preferably be rapid in order to minimize the length of the shutdown. To facilitate replacing a CRDM in the field, a standoff assembly connected to the distribution plate 22 to provide precise vertical placement of the CRDM 20 is also configured to provide electrical power and hydraulics to the CRDM 20 via connectors that require no action to effectuate the connection other than placement of the standoff assembly onto the distribution plate 22. After placement, the standoff is secured to the distribution plate by bolts or other fasteners. Additionally or alternatively, it is contemplated to rely upon the weight of the standoff assembly and CRDM to hold the assembly in place, or to use welds to secure the assembly. The illustrative distribution plate 22 is a single plate that contains the electrical and hydraulic lines and also is strong enough to provide support to the CRDMs and upper internals without reinforcement. In another embodiment, the distribution plate 22 may comprise a stack of two or more plates, for example a mid-hanger plate which provides structural strength and rigidity and an upper plate that contains electrical and/or hydraulic lines to the CRDMs via the standoff assembly. The motor/roller nut assembly of the CRDM is generally located in the middle of the lead screw's travel path. When the control rod is fully inserted into the core, the roller nut is holding the top of the lead screw, and, when the rod is at the top of the core, the roller nut is holding the bottom of the lead screw and most of the length of the lead screw extends upward above the motor/roller nut assembly. Hence the distribution plate 22 that supports the CRDM is positioned “below” the CRDM units and a relatively short distance above the reactor core. FIG. 2 illustrates the distribution plate 22 with one standoff assembly 24 mounted for illustration, though it should be understood that all openings 26 would have a standoff assembly (and accompanying CRDM) mounted in place during operation of the reactor. Each opening 26 allows a lead screw of a control rod to pass through and the periphery of the opening provides a connection site for a standoff assembly that supports the CRDM. The lead screw passes down through the CRDM, through the standoff assembly, and then through the opening 26. The distribution plate 22 has, either internally embedded within the plate or mounted to it, electrical power lines (e.g., electrical conductors) and hydraulic power lines to supply the CRDM with power and hydraulics. The illustrative openings 26 are asymmetric or keyed so that the CRDM can only be mounted in one orientation. As illustrated, there are 69 openings arranged in nine rows to form a grid, but more or fewer could be used depending on the number of control rods in the reactor. The distribution plate is circular to fit the interior of the reactor, with openings 28 to allow for flow through the plate. In some designs, not all openings may have CRDMs mounted to them or have associated fuel assemblies. One possible arrangement for the hydraulic and/or electrical power lines is shown in FIG. 3. The electrical power lines, shown as dashed lines 30, runs straight between the rows of openings 26 in the distribution plate 22. Because of the limited flexibility of typical cables compatible with the high temperature and caustic environment inside the pressure vessel, the power lines within the distribution plate 22 for delivering electrical and/or hydraulic power to the CRDMs should be straight or have gradual, large-radius turns To accommodate both electrical and hydraulic power lines, in one embodiment the hydraulic power lines (not shown) follow a similar pattern to that of the electrical lines 30. In another embodiment the hydraulic power lines follow a similar path, except that the pattern of hydraulic lines is rotated 90° from the electrical path. The hydraulic power lines and electrical power lines, if internal to the plate, are separated by depth in the plate. Alternatively, one or other can be disposed on a top or bottom surface of the plate 22, or they can be disposed on opposite top and bottom surfaces of the plate. FIG. 4 illustrates a small cutaway view of one connection site of the distribution plate 22 for connecting a CRDM to the distribution plate. The connection site includes an opening 26 for passing the lead screw of a single CRDM. Located around the opening 26 are apertures 40 to accept bolts (more generally, other securing or fastening features may be used) and electrical connectors 42 for delivering electrical power to the CRDM. The illustrative CRDM employs hydraulic power to operate the SCRAM mechanism, and accordingly there is also a hydraulic connector 44 to accept a hydraulic line connection. The opening 26 and its associated features 40, 42, 44 create a connection site to accept the CRDM/standoff assembly. Internal to the plate may be junction boxes to electrically connect the connection sites to the electrical power lines 30 running in between rows of connection sites. Similarly, the hydraulic connector 44 may connect to a common hydraulic line 32 running through the distribution plate perpendicular to the electrical power lines 30 and separated by depth. FIG. 5 illustrates a standoff 24 that suitably mates to opening 26 in the distribution plate 22. The standoff assembly has a cylindrical midsection with plates 45, 46 of larger cross-sectional area on either end of the midsection. The circular top plate 45 mates to and supports a CRDM 20. The square bottom plate 46 mates to the distribution plate 22. Although the illustrative bottom plate 46 is square, it may alternatively be round or have another shape. When the CRDM 20 and the top plate 45 of the standoff 24 are secured together they form a unitary CRDM/standoff assembly in which the bottom plate 46 is a flange for connecting the assembly to the distribution plate 22. Two bolt lead-ins 50 on diagonally opposite sides of the lower plate 46 mate to the apertures 40 of the distribution plate. The bolt lead-ins, being mainly for positioning the CRDM standoff, are the first component on the standoff to make contact with the distribution plate when the CRDM is being installed, ensuring proper alignment. Two electrical power connectors 52 on diagonally opposite corners of the bottom plate 46 mate to corresponding electrical power connectors 42 of the distribution plate 22. A hydraulic line connector 54 on the bottom plate 46 mates to the corresponding hydraulic power connector 44 of the distribution plate 22. A central bore 56 of the standoff 26 allows the lead screw to pass through. The connectors 42, 44 inside the distribution plate 22 optionally have compliance features, such as springs, belleville washers or the like, to ensure positive contact, and the opposing bolts that attach at lead-ins 50 serve as tensioning devices to ensure proper seating of both the CRDM electrical connectors and hydraulic connectors. Flow slots 58 permit primary coolant to flow through the standoff. FIG. 6 illustrates a perspective view focusing on the top plate 45 of the standoff 24. The top plate 45 of the standoff mates to the CRDM and is attached via bolt holes 62. Bolt holes 62 may be either threaded or unthreaded. The CRDM and standoff can be attached to each other and electrical connections 52 and hydraulic connection 54 made before the CRDM is installed so as to form a CRDM/standoff assembly having flange 46 for connecting the assembly with the connection site of the distribution plate 22. The bottom plate 46 of the standoff 24 is secured to the connection site via bolts passing through holes 50 and secured by nuts, threads in the bolt holes 40, or the like. FIG. 7 illustrates another suitable standoff 70, with a generally square upper mounting plate 71 for the CRDM. The upper mounting plate 71 for the CRDM includes a notch 76 to enable electrical access to the bottom of the CRDM, bolt holes 77 to attach the CRDM, and notches 78 at the corners of the plate to permit primary coolant flow. The lower mounting plate 72, which connects to the distribution plate 22, includes three electrical power connectors 73, a hydraulic power connector 74, and flow slots 75 to permit coolant flow. The standoff 70 may have more or fewer electrical connections depending on whether CRDM components share an electrical connection or have their own connection. FIG. 8 shows standoff 24 connected to a CRDM 20 to form a CRDM/standoff assembly that can be mounted to the distribution plate. CRDM electrical cabling 80 extends upward to conduct electrical power received at the electrical connectors 52 to the motor or other electrical component(s) of the CRDM 20. Similarly, a CRDM hydraulic line 82 extends upward to conduct hydraulic power received at hydraulic connector 54 to the hydraulic piston or other hydraulic component(s) of the CRDM 20 to maintain latching—removal of the hydraulic power instigates a SCRAM. The entire assembly including the CRDM and the standoff is then installed as a unit on a distribution plate, simplifying the installation process of a CRDM in the field. In one embodiment, the electrical cables 80 are mineral insulated cables (MI cables) which generally include one, two, three, or more copper conductors wrapped in a mineral insulation such as Magnesium Oxide which is in turn sheathed in a metal. The mineral insulation could also be aluminum oxide, ceramic, or another electrically insulating material that is robust in the nuclear reactor environment. MI cables are often sheathed in alloys containing copper, but copper would corrode and have a negative effect on reactor chemistry. Some contemplated sheathing metals include various steel alloys containing nickel and/or chromium, or a copper sheath with a protective nickel cladding. The electrical lines 30 in the distribution plate 22 (see FIG. 3) are also suitably MI cables, although other types of cabling can be used inside the distribution plate 22 if they are isolated by embedding in the plate. MI cables advantageously do not include plastic or other organic material and accordingly are well suited for use in the caustic high temperature environment inside the pressure vessel. The relatively rigid nature of the MI cables is also advantageous in that it helps ensure the integrity of the pre-assembled CRDM/standoff assembly during transport and installation. However, the rigidity of the MI cables limits their bending radius to relatively large radius turns, so that the MI cables inside the distribution plate 22 should be arranged as straight lines with only large-radius turns, e.g. as shown in FIG. 3. The large area of the distribution plate 22, which spans the inner diameter of the pressure vessel, facilitates a suitable arrangement of the MI cables inside the plate 22. Additionally, some types of MI cables are susceptible to degradation if the mineral insulation is exposed to water. Accordingly, the ends of the MI cables, e.g. at the coupling with the connector 52 in the standoff and the coupling of the power lines 30 with the electrical connectors 42 in the distribution plate 22, should be sealed against exposure to the primary coolant water. However, advantageously, the connectors 42, 52 themselves can be immersed in water. This makes installation, to be further described, readily performed even with the reactor core immersed in primary coolant. FIG. 9 is an overhead view of a standoff assembly with installed CRDM. This view would be looking down from the upper internals into the core when the CRDM and standoff assembly are mounted in the reactor. Connecting rod 90 is contained within lead screw 92 which is raised and lowered by the CRDM. Bolt holes 50 are visible at diagonally opposite corners. Cables can be seen running to electrical connectors 52 at the other pair of corners. A portion of the vertically extending CRDM hydraulic line 82 can be seen in end view. FIG. 10 shows a suitable configuration for the mating electrical connectors 42, 52 of the distribution plate and CRDM/standoff assembly flange 46, respectively. The female electrical connector 52 of the standoff assembly 24 lowers onto and covers the male electrical connector 42 of the distribution plate. The connectors 42, 52 preferably include glands or other features to prevent ingress of water to the mineral insulation of the MI cables 30, 80 at the junctions of these cables with the respective connectors 42, 52. In this way, the connectors 42, 52 can be mated underwater without exposing the metal insulation, so as to facilitate installing the CRDM/standoff assembly at the connection site of the distribution plate 22 while keeping the reactor core and the distribution plate 22 submerged in primary coolant. To ensure a good electrical connection, the connection between connectors 42, 52 can be purged to evacuate any trapped water. Alternatively, the electrical connectors could be mated and not purged, albeit typically with some increased resistance due to wet connectors. FIG. 11 shows a suitable hydraulic interface from standoff assembly 24 to distribution plate 22. An electrical connector 52 as already described with reference to FIG. 10 is also shown. The female hydraulic connector 54 of the standoff assembly mates to the male hydraulic connector 110. The female hydraulic connector 54 is a socket that is machined directly into the bottom of the lower plate 72 of the standoff assembly 24. The top of the hydraulic connector 54 has a nipple to allow the hydraulic line 82 to be connected to the standoff assembly 22. The hydraulic line then runs up the CRDM to a piston assembly (not shown) which latches the lead screw. The hydraulic connectors 54, 110 optionally have compliance features, such as springs, belleville washers or the like, to ensure positive contact. A continuous flow of primary coolant is used as hydraulic fluid to maintain the CRDM latched during operation, so some leakage from the hydraulic connector into the pressure vessel is acceptable. In view of this, in some embodiments the mating of the hydraulic power connector of the CRDM 20 with the corresponding hydraulic power connector of the connection site of the distribution plate 22 forms a leaky hydraulic connection. Accordingly, a sufficient sealing force for the hydraulic connection is provided by the weight of the CRDM/standoff assembly and/or the force imparted by the hold-down bolts that pass through the bolt lead-ins 50 of the standoff assembly and bolt holes 40 of the distribution plate. FIG. 12 diagrammatically illustrates a method of connecting a CRDM to a standoff to form a preassembled CRDM/standoff assembly and then connecting the CRDM/standoff assembly to the distribution plate. In step S1210, the method starts. In step S1220, the CRDM is bolted to the standoff assembly by a plurality of bolts. In step S1230, the hydraulic cable is connected to the hydraulic connector of the standoff plate and the electrical cable(s) are connected the electrical connection(s). In step S 1240, the standoff plate, with CRDM bolted on top of it, is lowered onto the distribution plate, with the bolt holes 50 making contact first to ensure proper alignment of the standoff assembly and CRDM. In step S 1250, the hold-down bolts are installed and torqued to attach the standoff assembly to the distribution plate and to ensure positive contact in the hydraulic and electrical connectors. At step S1260, the electrical connectors are optionally purged. At step S1270, the method ends. FIG. 13 illustrates a method of removing a CRDM from a distribution plate. In step S1310, the method starts. In step S1320, the hold-down bolts are removed. In step S1330, the CRDM and connected standoff assembly are lifted away from the distribution plate. In step S1340, the CRDM is optionally removed from the standoff assembly for repair or replacement. In step S1350, the method ends. The disclosed approaches advantageously improve the installation and servicing of powered internal mechanical reactor components (e.g., the illustrative CRDM/standoff assembly) by replacing conventional in-field installation procedures including on-site routing and installation of power lines (e.g. MI cables or hydraulic lines) and connection of each power line with the powered internal mechanical reactor component with a simple “plug-and-play” installation in which the power lines are integrated with the support plate and power connections are automatically made when the powered internal mechanical reactor component is mounted onto its support plate. The disclosed approaches leverage the fact that most powered internal mechanical reactor components are conventionally mounted on a support plate in order to provide sufficient structural support and to enable efficient removal for servicing (e.g., a welded mount complicates removal for servicing). By modifying the support plate to also serve as a power distribution plate with built-in connectors that mate with mating connectors of the powered internal mechanical reactor component during mounting of the latter, most of the installation complexity is shifted away from the power plant and to the reactor manufacturing site(s). The example of FIGS. 1-13 is merely illustrative, and numerous variations are contemplated. For example, the CRDM/standoff assembly can be replaced by a CRDM with an integral mounting flange, that is, the standoff can be integrally formed with the CRDM as a unitary element (variant not shown). With reference to FIGS. 14 and 15, as another illustrative example the disclosed approaches are applied to internal reactor coolant pumps (RCPs) 1400, such as are disclosed in Thome et al., U.S. Pub. No. 2010/0316181 A1 published Dec. 16, 2010 which is incorporated herein by reference in its entirety. For placement of the internal RCPs 1400 in the cold leg (i.e. the downcomer annulus), the RCPs 1400 are envisioned to be mounted on an annular pump plate 1402 disposed in the downcomer annulus. The pump plate 1402 serves as structural support for the RCPs 1400 and also as a pressure divider to separate the upper suction volume and the lower discharge volume. In the illustrative embodiment there are eight connection sites with six of these shown in FIG. 14 as containing RCPs 1400, and the remaining two being unused to illustrate the connection sites. The pump plate 1402 is modified to include MI cables 1404, 1405 disposed in or on the pump plate 1402. The annular shape of the pump plate 1402 precludes long straight runs of MI cable; however, the illustrative MI cables 1404, 1405 are oriented circumferentially with a large bend radius comparable with (half of) the inner diameter of the pressure vessel 11. Bolt apertures 1440 and electrical connectors 1442 are analogous to bolt apertures 40 and electrical connectors 42 of the illustrative CRDM embodiment, respectively. The opening 26 of the connection site of distribution plate 22 translates in the pump plate 1402 to be a generally circular opening 1426 (optionally keyed by a suitable keying feature, not shown) through which the RCPs 1400 pump primary coolant downward. As yet another contemplated modification, it will be appreciated that the female connector can be located in the supporting power distribution plate while the male connector can be located in the flange, standoff or other mounting feature of the internal mechanical reactor component. The preferred embodiments have been illustrated and described. Obviously, modifications and alterations will occur to others upon reading and understanding the preceding detailed description. It is intended that the invention be construed as including all such modifications and alterations insofar as they come within the scope of the appended claims or the equivalents thereof.
	abstract	A method of storing high level radioactive waste, and specifically a method of adjusting or controlling the temperature of ventilation air flowing through a storage cavity of a ventilated system. The method includes positioning a metal canister containing high level radioactive waste in a storage cavity of the ventilated system. The ventilated system includes a cask body, a cask lid, a plurality of inlet ducts, and at least one outlet duct so that ventilation air can flow from atmosphere into the storage cavity where it is heated and then back out to the atmosphere. The method includes progressively reducing a cross-sectional area of one or more of the inlet ducts and/or the outlet duct over time so that a rate at which the ventilation air is heated within the storage cavity is maintained above a predetermined threshold to mitigate the risk of stress corrosion cracking in the metal canister.
	summary
051749475	claims	1. A pellet turning apparatus for facilitating surface inspection of nuclear fuel pellets, comprising: (a) a pellet turner assembly capable of supporting a plurality of pellets and having a pellet loading end, said pellet turner assembly being operable for producing simultaneous turning of the pellets in situ to permit visual inspection of the entire circumferential surfaces thereof; and (b) an interface assembly coupled to said pellet turner assembly and being operable for positioning a pellet transfer end of a pellet supply tray relative to said loading end of said pellet turner assembly to facilitate transfer of uninspected pellets from the pellet supply tray to said pellet turner assembly and of inspected pellets from said pellet turner assembly to the pellet supply tray; (c) said pellet turner assembly including a rake engageable with the pellets and for use by an operator to transfer pellets between the pellet supply tray and said pellet turner assembly. a tray elevating mechanism mounted to said pellet loading end of said pellet turner assembly for engaging the transfer end of the pellet supply tray; and an actuating cam mechanism coupled to said pellet elevating mechanism and being operable for moving said elevating mechanism to position the pellet transfer end of the supply tray above or below the loading end of said pellet turner assembly. a pellet turner deck; and a swivel mechanism supporting said pellet turner deck. a base frame mounted on said swivel mechanism; a plurality of elongated laterally-spaced members extending between opposite ends of said base frame and defining channels for receiving the pellets in rows thereof between said members; and means for supporting said elongated channel-defining members at their opposite ends in spaced relation above said base frame. a pellet rolling plate supported on said base frame below said pellet channel-defining members for supporting pellets between said pellet channel-defining members; and an actuating mechanism coupled to said pellet rolling plate and being operable to cause oscillatory movement of said plate relative to said channel-defining members for producing turning of the pellets in situ within said channels. said loading end of said pellet turner assembly is defined at one end of said pellet turner deck; and said interface assembly is mounted at said one end of said pellet turner deck. a tray elevating mechanism mounted to said one end of said pellet turner deck for engaging the transfer end of the pellet supply tray; and an actuating cam mechanism coupled to said pellet elevating mechanism and being operable for moving said elevating mechanism to position the pellet transfer end of the supply tray above or below the loading end of said pellet turner assembly. (a) a pellet turner assembly capable of supporting a plurality of pellets and having a pellet loading end, said pellet turner assembly being operable for producing simultaneous turning of the pellets in situ to permit visual inspection of the entire circumferential surfaces thereof, said pellet turner assembly being further operable for placing the pellets simultaneously in a tilted position to permit visual inspection of opposite edges thereof; and (b) an interface assembly coupled to said pellet turner assembly and being operable for disposing a pellet transfer end of a pellet supply tray at a desired elevation above or below said loading end of said pellet turner assembly to facilitate transfer of uninspected pellets from the pellet supply tray to said pellet turner assembly and of inspected pellets from said pellet turner assembly to the pellet supply tray; (c) said pellet turner assembly including a pellet turner deck for supporting pellets in side-by-side spaced rows; and a swivel mechanism supporting said pellet turner deck and being operable for tilting said pellet turner deck to produce simultaneous tilting of the pellets to permit visual inspection of opposite edges thereof. a base frame mounted on said swivel mechanism; a plurality of elongated laterally-spaced members extending between opposite ends of said base frame and defining channels for receiving and arranging the pellets in rows thereof between said members; and means for supporting said elongated channel-defining members at their opposite ends in spaced relation above said base frame. a pellet rolling plate supported on said base frame below said pellet channel-defining members for supporting pellets between said pellet channel%defining members; and an actuating mechanism coupled to said pellet rolling plate and being operable to cause oscillatory movement of said plate relative to said channel-defining members for producing turning of the pellets in situ within said channels. said loading end of said pellet turner assembly is defined at one end of said pellet turner deck; and said interface assembly is mounted at said one end of said pellet turner deck. a tray elevating mechanism mounted to said one end of said pellet turner deck for engaging the transfer end of the pellet supply tray; and an actuating cam mechanism coupled to said pellet elevating mechanism and being operable for moving said elevating mechanism to position the pellet transfer end of the supply tray above or below the loading end of said pellet turner assembly. (a) a pellet turner assembly capable of supporting a plurality of pellets and having a pellet loading end, said pellet turner assembly being operable for producing simultaneous turning of the pellets in situ to permit visual inspection of the entire circumferential surfaces thereof; and (b) an interface assembly coupled to said pellet turner assembly and being operable for positioning a pellet transfer end of a pellet supply tray relative to said loading end of said pellet turner assembly to facilitate transfer of uninspected pellets from the pellet supply tray to said pellet turner assembly and of inspected pellets from said pellet turner assembly to the pellet supply tray; (c) said pellet turner assembly including (a) a pellet turner assembly capable of supporting a plurality of pellets and having a pellet loading end, said pellet turner assembly being operable for producing simultaneous turning of the pellets in situ to permit visual inspection of the entire circumferential surfaces thereof, said pellet turner assembly being further operable for placing the pellets simultaneously in a tilted position to permit visual inspection of opposite edges thereof; and (b) an interface assembly coupled to said pellet turner assembly and being operable for disposing a pellet transfer end of a pellet supply tray at a desired elevation above or below said loading end of said pellet turner assembly to facilitate transfer of uninspected pellets from the pellet supply tray to said pellet turner assembly and of inspected pellets from said pellet turner assembly to the pellet supply tray; (c) said pellet turner assembly including 2. The apparatus as recited in claim 1, further comprising: 3. The apparatus as recited in claim 1, wherein said interface assembly is operable for disposing said pellet transfer end of the pellet supply tray at a desired elevation above or below said pellet loading end of said pellet turner assembly. 4. The apparatus as recited in claim 3, wherein said interface assembly includes: 5. The apparatus as recited in claim 4, wherein said tray elevating mechanism includes a plurality of engaging elements capable of coupling with the pellet transfer end of the pellet supply tray, said actuating cam mechanism being operable for raising or lowering of said engaging elements upon movement of said elevating mechanism and thereby to raise or lower the pellet transfer end of the supply tray respectively above or below said loading end of said pellet turner assembly to facilitate transferring of pellets between said pellet turner assembly and the pellet supply tray. 6. The apparatus as recited in claim 1, wherein said pellet turner assembly includes: 7. The apparatus as recited in claim 6, wherein said pellet turner deck includes: 8. The apparatus as recited in claim 7, wherein said pellet turner deck further includes: 9. The apparatus as recited in claim 6, wherein: 10. The apparatus as recited in claim 9, wherein said interface assembly is operable for disposing said pellet transfer end of the pellet supply tray at a desired elevation above or below said one end of said pellet turner deck. 11. The apparatus as recited in claim 10, wherein said interface assembly includes: 12. The apparatus as recited in claim 11, wherein said tray elevating mechanism includes a plurality of engaging elements capable of coupling with the pellet transfer end of the pellet supply tray, said actuating cam mechanism being operable for raising or lowering of said engaging elements upon movement of said elevating mechanism and thereby to raise or lower the pellet transfer end of the supply tray respectively above or below said loading end of said pellet turner assembly to facilitate transferring of pellets between said pellet turner assembly and the pellet supply tray. 13. A pellet turning apparatus for facilitating surface inspection of nuclear fuel pellets, comprising: 14. The apparatus as recited in claim 13, wherein said pellet turner assembly includes: 15. The apparatus as recited in claim 14, wherein said pellet turner deck includes: 16. The apparatus as recited in claim 15, wherein said pellet turner deck further includes: 17. The apparatus as recited in claim 14, wherein: 18. The apparatus as recited in claim 17, wherein said interface assembly is operable for disposing said pellet transfer end of the pellet supply tray at a desired elevation above or below said one end of said pellet turner deck. 19. The apparatus as recited in claim 18, wherein said interface assembly includes: 20. The apparatus as recited in claim 19, wherein said tray elevating mechanism includes a plurality of engaging elements capable of coupling with the pellet transfer end of the pellet supply tray, said actuating cam mechanism being operable for raising or lowering of said engaging elements upon movement of said elevating mechanism and thereby to raise or lower the pellet transfer end of the supply tray respectively above or below said loading end of said pellet turner assembly to facilitate transferring of pellets between said pellet turner assembly and the pellet supply tray. 21. A pellet turning apparatus for facilitating surface inspection of nuclear fuel pellets, comprising: 22. A pellet turning apparatus for facilitating surface inspection of nuclear fuel pellets, comprising:
	description	This application is based upon and claims the benefit of priority from Japanese Patent Application No. 2009-194829, filed on Aug. 25, 2009, the entire contents of which are expressly incorporated herein by reference. A. Field Embodiments described herein relate generally to radiation therapy equipment having a multi-divided irradiation collimator that can accurately determine an irradiation field of radiation as similar to a configuration of a desired region on an object. B. Background Typically, the radiation therapy equipment has an irradiation head unit for irradiating radiation on an object. In the irradiation head unit, an irradiation collimator (i.e., radiation collimator) is installed to determine an irradiation field on a desired treatment target region, such as a malignant tumor region of an object for preventing a radiation hazard from unnecessary radiation regions on the object. To reduce the radiation hazard as much as possible, it is required for the irradiation collimator to determine the irradiation field as accurately as possible to a configuration of the target treatment region. Usually, the irradiation collimator is comprised of an upper pair of collimator blocks provided at a near side of a radiation source along the radiation axis and a lower pair of collimator blocks provided at a lower position than the upper pair of collimator blocks along the radiation axis. The lower pair of collimator blocks is positioned so as to orthogonally cross the upper pair of collimator blocks. The pair of upper collimator blocks is provided so as to face with each other with centering an irradiation axis of the radiation. The pair of upper collimator blocks is driven so as to approach or get away from each other along an arc shaped tracking direction (X-direction). The arc is a portion of a circle with centering the radiation source. The pair of lower collimator blocks is also provided so as to face with each other with centering an irradiation axis of the radiation. The pair of lower collimator blocks is driven so as to approach or get away from each other along an orthogonal arc shaped tracking direction (Y-direction) to the X-direction for tracking by the upper collimator blocks. Each of the lower collimator blocks is comprised of a multi-divided collimator block that is constructed by a plurality of closely attached leaf plates. Each of the plurality of leaf plates has an arc shaped tracking surface facing the radiation irradiation axis and a toothed configuration of a screw is provided on the tracking surface in order to engage with a driving gear. The driving gear is fixed to a tip portion a rotation shaft. The shaft is rotated by a motor of a driving source through a driving power transmitting mechanism such as worm gears. A detection unit, such as a potentiometer or an encoder is provided to detect a driven amount of the gears. Based on the data detected by the detection unit, each of the leaf plates in the lower collimator blocks is independently driven to a desired position. Thus, in a conventional radiation therapy equipment, an irradiation field is formed so as to be approximate to an irregular shape of a target treatment region by moving the pair of upper collimator blocks in the X direction and each of the plurality of leaf plates in each of the pair of lower collimator blocks in the Y direction being orthogonal to the X direction. However, the conventional irradiation collimator needs to use the driving power transmitting mechanism to rotate a driving gear through a shaft by transmitting a driving power of a motor to independently move each of the plurality of leaf plates in the lower collimator block. Since the conventional driving power transmitting mechanism is constructed by combining various types of toothed wheels, backlash due to the respective toothed wheels are accumulated, it has difficult to accurately control to position each of the leaf plates in the irradiation collimator. Further, in the conventional irradiation collimator, the driving power transmitting mechanism needs to be provided so as that the motor must be positioned in parallel or at a right angle to the rotation shaft for each of the leaf plates. Due to such a limitation, it has been difficult to increase the number of the shaft driving power transmitting mechanism to increase the number of the leaf plates in a limited space of the irradiation head unit. Recently, to protect against radiation hazards, a requirement for setting an irradiation field as accurately close to a configuration of a target treatment region is strongly increased. To set the irradiation field as closely as possible to an irregular shape of the target treatment region, the number of the leaf plates in the lower collimator blocks needs to be increased with decreasing each thickness of the leaf plates. However, as mentioned above, it has been difficult to increase the number of the shaft driving power transmitting mechanisms to increase the number of the leaf plates in a limited space of the irradiation head unit. An embodiment of the present invention addresses these and other problems and drawbacks and provides an irradiation collimator installing in a radiation therapy equipment that can perform a high accuracy radiation therapy. While a number of the driving power transmission mechanisms are increased with increasing a number of the leaf plates, they can be freely installed in a limited space of the irradiation unit. According to an embodiment, a plurality of constant force springs with no change in torque regardless of a length of stroke are used for driving each of the plurality of leaf plates. The plurality of constant force springs are coaxially supported on a pair of rotation shafts. Each rotation shaft is fixed to each of a plurality of driving gears. Each of the driving gears is engaged to a toothed configuration provided on each tracking surface of the leaf plates. As a result, a constant force is applied to each of the leaf plates so as to move in a closing direction. While a constant force is applied to each of the leaf plates is constantly in a closing direction by the constant force spring, each of the leaf plates is moved in an opening direction by rewinding up a wire connected to each of plates through a pulley. By using a plurality of constant force springs, the conventional backlashes due to the usage of a plurality of driving gears can be substantially avoided and an accurate irradiation field can be precisely set up by increasing the number of the leaf plates. Further, an increased number of transmission mechanisms for moving each of the increased number of leaf plates also can be freely installed in a limited space of the irradiation unit. An embodiment of the radiation therapy equipment includes a radiation collimator unit for shielding radiation from a radiation source so as to limit an irradiation field of the radiations onto a target treating region. The radiation collimator unit is comprised of a pair of movable collimator blocks, each block including a plurality of movable leaf plates, each leaf plate having a toothed configuration on an arc shaped tracking surface of the leaf plate, a plurality of driving gears configured to independently drive each of the plurality of leaf plates by engaging with the toothed configuration provided on each tracking surface of the plurality of leaf plates, a plurality of constant force spring units respectively coupled to each of the plurality of driving gears configured to constantly apply force to each of the plurality of driving gears so as to independently move each of the plurality of leaf plates in closing directions, a plurality of wires connected to each of the plurality of leaf plates, and a plurality of roll-up units configured to independently roll-up each of the plurality of wires so as to independently move each of the leaf plates in an opening direction against each of the closing forces of the plurality of constant force spring units. Radiation therapy equipment according to the embodiment can safely protect an object even when the wire connected to the movable leaf plate is accidentally cut because the leaf plate immediately closes the irradiation aperture by the closing force due to the constant force spring unit. Reference will first be made to the entire construction of a radiation therapy equipment according to an exemplary embodiment of the present invention. FIG. 1 illustrates an entire construction of radiation therapy equipment according to this embodiment. Radiation therapy equipment 100 is comprised of a radiation irradiation section 10 for irradiating radiation emitted from a radiation source on an object P, a treatment table section 20 for positioning of an irradiation region by moving a top plate 22 supporting the object P, a control section 30 for totally controlling the radiation irradiation section 10 and the treatment table section 20. The radiation irradiation section 10 includes a fixed gantry 11 provided on a floor of the radiation therapy room, a rotation gantry 12 provided on the fixed gantry 11 so as to rotate around a horizontal rotation axis H of the fixed gantry 11, an irradiation head unit 13 provided on a head portion of the movable gantry 12 for irradiating ray beams on an affected region of the object P, and a rotatable irradiation collimator 14 installed in the irradiation head unit 13. As shown in FIG. 1, the irradiation collimator 14 may rotate around an irradiation axis i of the irradiation head unit 13. The horizontal rotation axis H of the rotation gantry 12 crosses the irradiation axis I of the ray-beam irradiation head unit 13 at an iso-center IC. The rotation gantry 12 can be moved in correspondence to various irradiation manners, such as a rotation irradiation, a pendulum irradiation and an intermittent irradiation. The treatment table section 20 can be rotated on the floor in an arrow direction G at a prescribed angle along an arc circle with centering the iso-center IC. The treatment table section 20 includes an upper mechanism 21 for supporting the top plate 22, a lifting mechanism 23 for lifting the upper mechanism 21, and a lower mechanism 24 for supporting the lifting mechanism 23. The lifting mechanism 23 moves up and down the upper mechanism 21 and the top plate 22 in an arrow D direction. The lower mechanism 24 rotates the lifting mechanism 23 around an axis passing the iso-center IC at a position F that has a distance L in a horizontal (D) direction from the IC axis. In addition to the rotation of the lifting mechanism 23, the upper mechanism 21 and the top plate 22 also rotate at a prescribed angle in the horizontal direction. The control section 30 includes an operation unit (not shown). Through the operation unit, a medical staff sets up a position of the top plate 22 supporting an object P and an irradiation field determined by the irradiation collimator 14. In the irradiation unit 13, an irradiation collimator 14 is installed so as to rotate around the irradiation axis I. The irradiation collimator 14 limits radiation so as to irradiate radiation just on a desired target treatment region, such as a malignant tumor, while avoiding unnecessary irradiation on normal tissues. The irradiation collimator 14 is made of a heavy metal, such as tungsten, for prohibiting radiation onto unnecessary regions, i.e., normal tissues. FIG. 2 is a cross-sectional view of the irradiation collimator 14 along Y axis direction. As illustrated in figure, the irradiation collimator 14 is comprised of a pair of upper collimator blocks 140A and 140B facing the irradiation axis I and a pair of lower collimator blocks 141A and 141B. The pair of upper collimator blocks 140A and 140B is positioned at a radiation source S side. The pair of lower collimator blocks 141A and 141B is provided under the pair of upper collimator blocks 140A and 140B and arranged so as to be orthogonal to the arranged direction of the upper collimator blocks. Toothed configurations 140A1 and 140B1 are respectively provided on each under surface of the upper collimator blocks 140A and 140B so as to engage to a pair of driving gears 140A2 and 140B2. A pair of driving units 142A and 142B drives the pair of driving gears 140A2 and 140B2 to approach or move apart the pair of upper collimator blocks 140A and 140B in the arrow X direction along an arc shaped tracking surface with each other. A pair of lower collimator blocks 141A and 141B is provided under the pair of upper collimator blocks 140A and 140B so as to face the irradiation axis I with each other. The lower collimator blocks 141A and 141B are constructed by a plurality of leaf plates 141Am (m=1−n) and 141Bm (m=1−n), respectively. FIG. 3 is a side view of the irradiation collimator 14 viewing an orthogonal direction of FIG. 2. The pair of lower collimator blocks 141A and 141B is orthogonally provided under the pair of upper collimator blocks 140A and 140B facing the irradiation axis I with each other. Toothed configurations are respectively provided on each under surface of the lower collimator blocks 141A and 141B so as to engage to a pair of driving gears 150A and 150B. A pair of driving units 143A and 143B drives the pair of driving gears 150A and 150B to approach or move apart the pair of lower collimator blocks 141A and 141B in the arrow Y direction along an arc shaped tracking surface with each other. As illustrated in FIG. 2, each of the pair of lower collimator blocks 141A and 141B is a multi-divided collimator block that is comprised of a plurality of leaf plates. Thus, the collimator block 141A is constructed by closely adjoining a plurality of leaf plates 141 Am (m=1˜n). And the collimator block 141B is constructed by closely adjoining a plurality of leaf plates 141Bm (m=1˜n). FIG. 4 is a plane view of the pair of multi-divided irradiation collimator blocks. Toothed configurations are provided on each tracking surface of the leaf plates constructing the lower collimator blocks 141A and 141B. The toothed configurations on each of the leaf plates are respectively engaged to a plurality of driving gears. Each of the driving gears is respectively driven by a plurality of drive units 143A1 through 143An, and 143B1 through 143Bn so as to respectively move the plurality of leaf plates 141A1 through 141An and 141B1 through 141Bn in desired positions. Thus, the irradiation collimator 14 moves the pair of upper collimator blocks 140A and 140B so as to approach to or secede from each other in the X direction passing through a center axis of the radiation source S. Further the irradiation collimator 14 moves each of the leaf plates 141A1 through 141An and 141B1 through 141Bn in the pair of lower multi-divided irradiation collimator blocks 141A and 141B so as to approach to or move away from each other in the Y direction passing through a center axis of the radiation source S. By moving both upper collimator blocks 140A and 140B and each of the leaf plates in the lower multi-divided irradiation collimator blocks, a desired irradiation field U (FIG. 5) is formed. FIG. 5 depicts the irradiation field U formed on an irregular shape T of a target treatment region so as closely as possible by the irradiation collimator 14. To make the shape of the irradiation field U approximate to the irregular shape T of the target treatment region, it is required to accurately move each of the leaf plates 141A1 through 141An and 141B1 through 141Bn in the lower collimator blocks, independently, so as to coincide with the irregular shape T. To accurately move each of the leaf plates 141A1 through 141An and 141B1 through 141Bn, backlashes need to be avoided due to the driving mechanism including each of the driving gears that respectively engage with each toothed configuration formed on an arc surface of the leaf plate and each position of the leaf plate needs to be accurately detected. FIG. 6 illustrates the driving mechanism according to the present embodiment for driving a leaf plate 141m (m=1˜n) of the multi-divided irradiation collimator block. A left-side edge surface 141mb of the leaf plate 141m constructs an edge surface of the irradiation field as illustrated in FIG. 4. Thus, when the left-side edge surface 141mb of the leaf plate 141m is moved in the left direction of the drawing and a right-side edge surface of an opposite side leaf plate (not shown) facing the leaf plate 141m moves the right direction, an irradiation field formed by the pair of collimator leaf plates 141m is closed. On the contrary, each of the pair of leaf plates 141m moves in an opposite direction with each other, the irradiation field is widely opened. As illustrated in FIG. 6, each of the leaf plates 141m has an arc shaped tracking surface, and toothed configurations (not shown) are formed on an outer side surface 141ma of the leaf plates 141m. A driving gear 150 is engaged with the toothed configurations provided the outer side surface 141ma. The driving gear 150 is coaxially fixed to either one drum 161 of the constant force spring 160 for constantly tensioning the leaf plate 141m in a closing direction through the driving gear 150. FIG. 7A illustrates a construction of the constant force spring 160 for constantly tensioning the driving gear 150. The constant force spring 160 has a characteristic feature that a constant torque (output power) is kept regardless the stroke (number of turns). As illustrated in FIG. 7A, the constant force spring 160 rolls up a belt-like spring 163 on a first drum 161 and also reversely rolls up the belt-like spring 163 on a second drum 162. When the belt-like spring 163 is rolled up on the second drum 162, a constant rotation torque toward the first drum 161 is tensioned regardless the rolled up length (stroke) of the belt-like spring 163. FIG. 8 shows the characteristic features of the constant force spring. Wherein, the horizontal axis shows the stroke of the roll upspring 163, and the vertical axis shows the rotation torque. In a central hollow portion of the drum 161 of the constant force spring 160, a first shaft 171 is rotatably inserted. Similarly, a second shaft 172 is rotatably inserted into a central hollow portion of the second drum 162 of the constant force spring 160. The second rotation shaft 172 is provided in parallel to the first rotation shaft 171. The first rotation shaft 171 is fixed to the driving gear 150. FIG. 7B is a cross-sectional view of the first drum of the constant force spring along A-A direction shown in FIG. 7A. As illustrated in FIG. 7B, the first drum 161 of the constant force spring is constructed by rolling up a belt-like spring. One edge portion of the belt-like spring is coaxially fixed to a hollow portion of the driving gear 150. A first shaft 171 is inserted through a central hollow portion of the first drum 161 and rotatably supported through bearings 180. Similarly, a second shaft 172 is rotatably inserted into a central hollow portion of the second drum 162 of the constant force spring 160. Thus, one constant force spring 160 is supported by a pair of rotation shafts 171 and 172, and the leaf plate 141m is constantly tensioned in a closing direction of the irradiation field by the driving gear 150 foxed to either one drum of the constant force spring 160. As illustrated in FIG. 6, a right-side edge surface of the leaf plate 141m is connected to a wire 151. The wire 151 may be a piano wire. The wire 151 is rolled up by an axis of a motor 153 through at least one pulley 152. When the wire 151 is rolled up by rotations of the motor 153, the leaf plate 141m is moved in an opening direction (the right direction of the drawing) while a force is constantly applied to the leaf plate 141m in the closing direction by the driving gear 150 coaxially connected to the constant force spring 160. Thus, the leaf plate 141m is moved in the opening direction so as to position of a desired irradiation field by rolling up the wire 151 through the motor. When the leaf plate 141m is moved in the opening direction, the driving gear 150 connected to the constant force spring 160 applies a constant force to the leaf plate 141m in the closing direction by engaging with the toothed configuration of the leaf plate 141m. By doing so, backlashes due to engagement with the driving gear 150 are substantially eliminated. The moving amount of the leaf plate 141m in the opening direction is controlled by a roll-up amount of the motor 153. Thus, the control section 30 (FIG. 1) controls the roll-up operation of the motor 153 so as to set up the irradiation field U of the leaf plate 141m at a position in accordance with the shape T of the target treatment region. As illustrated in FIG. 4, each of the pair of lower collimator blocks 141A and 141B is respectively constructed by a plurality of leaf plates 141A1 through 141An and 141B1 through 141Bn. To independently move all of the leaf plates in the lower collimator block, a plurality of pairs of rotation shafts is provided along the moving direction of the lower collimator blocks for supporting the same number of driving mechanisms to the plurality of constant force springs 160. Each pair of rotation shafts respectively supports a different plural number of constant force springs in the X direction. Thus, in this embodiment, four (4) pairs of rotation shafts are provided in the Y direction, where a first pair of rotation shaft supports eleven (11) constant force springs, a second pair of rotation shaft supports twelve (12) constant force springs, and third and fourth pairs of rotation shafts respectively support nine (9) constant force springs as illustrated in FIG. 11. FIG. 9 is a side view of an example of driving mechanism of the either one of the lower collimator blocks. In the embodiment, four pairs of rotation shafts are provided for driving each of leaf plate 141m. For instance, one of a plurality of driving gears 150a fixed on one of a plurality of constant force springs 160a that is supported on a first pair of rotation shafts 171a and 172a is engaged with a first leaf plate 1411, and another one of the plurality of driving gears supported by the same first pair of rotation shafts 171a and 172a is engaged, for example, with each toothed configuration of a fifth leaf plate 1415 and a tenth leaf plate 14110 (either not shown). Similarly, the driving gears 150b of a plurality of constant force spring 160b supported by the second pair of rotation shafts 171b and 172b engaged with the toothed configuration of the second leaf plate 1412, and another driving gears supported by the second rotation shafts are engaged with the toothed configurations of, for instance, the sixth leaf plate 1416 and the eleventh leaf plate 14111. Similarly, the driving gears 150c of the plurality of constant force spring 160c supported by the third pair of rotation shafts 171c and 172c engage with the toothed configurations of the third leaf plate 1413, the seventh leaf plate 1417 and the twelfth leaf plate 14112. The driving gears 150d of the plurality of constant force spring 160d supported by the fourth pair rotation shafts 171d and 172d engage with the toothed configurations of the fourth leaf plate 1414, the eighth leaf plate 1418 and the thirteenth leaf plate 14113. By such a construction, a large number of driving gears can respectively engage with the large number of constant force springs by supporting the large number of constant force springs with a small number of rotation shaft pairs arranged along a tracking surface direction (X direction) of the lower irradiation collimator block. Consequently, it becomes possible to independently move an increased large number of leaf plates 141m in a limited space of the irradiation unit 13. FIG. 10 is a perspective view of the leaf plate driving mechanism in FIG. 9. FIG. 11 is an under perspective view of the leaf plate driving mechanism in FIG. 10. Only one block of the lower collimator blocks 141A and 141B is depicted in FIGS. 10 and 11. In the present embodiment, different number of constant force springs is respectively supported on each of four pairs of rotation shafts 171a and 172a through 171d and 172d. For instance, each of the driving gears 1501, 1502, - - - , 15011 fixed to a plurality of constant force springs 1601, 1602, - - - , 16011 supported by the first pair of rotation shafts 171a and 172a is engaged with the toothed configurations provided on each of the leaf plates 1411, 1412, - - - , 14111 for driving each of the leaf plates. In the embodiment, eleven of constant force springs 1601, 1602, - - - , 16011 are supported by the second pair of rotation shafts 171b and 172b along the X direction. the third pair of rotation shafts 171c and 172c and on the fourth pair of rotation shafts 171d and 172d, nine of the constant force springs 1601, 1602, - - - , 1609 are supported in the X direction. Thus, by supporting an appropriate number of constant force springs 160 on one pair of rotation shafts and also by arranging the plural number of constant force springs supported on different pairs of rotation shaft in staggered position with each other in the X direction, it becomes possible to install an increased number of leaf plates 141m in a limited space of the irradiation head unit. The opening and closing operations of the leaf plate 141m for forming an irradiation field U as approximately to a configuration T of the target treatment region (FIG. 5) are performed by rolling up each wire through each motor 153 under controlling of the control section 30. FIG. 12 illustrates a construction of the position detection unit 210 for a leaf plate 141m. FIG. 13 illustrates a partial section view of an outer edge portion 141ma of the leaf plate 141m. The toothed configuration 141mb (not shown) is formed on an arc shaped outer edge surface 141ma of a leaf plate 141m. A driving gear 150 fixed to the first drum 161 of each constant force spring 160 is engaged with the toothed configuration 141mb. As illustrated in FIG. 13, the outer edge surface 141ma of the leaf plate includes a toothed configuration portion 141mb and a non-toothed configuration portion 141mb in a width direction of the plate. On the non-toothed configuration portion 141mb, a light reflecting pattern 200 is fixed along the moving direction (Y direction) of the leaf plate 141m. FIG. 14 is an exemplary light reflecting pattern 200. As illustrated in FIG. 12, a light is emitted from a light emitting unit 211 to a fixed point Fx of the pattern 200 fixed on the outer edge surface 141ma of the leaf plate 141m. An imaging unit 212 provided near the leaf plate 141m acquires an image of the fixed point by imaging over a region including the fixed point Fx on the pattern 200. The light emitting unit 211 may be constructed by, for instance, light emitting diodes. The imaging unit 212 may be constructed by, for instance, a CCD camera. Output signal from the imaging unit 212 is supplied to a moving amount calculation unit 213. The moving amount calculation unit 213 may be provided in the control section 30. FIG. 14 is an exemplary light reflecting pattern 200. The pattern 200 is, as an example, a square region of 0.5 mm in four directions is divided into sixteen small squares in a net-like structure. By blacking out each corner of the small squares, special patterns 200a1 through 200a5 are formed. The special pattern is, of course, formed in other techniques than the coloring. The light emitting unit 211 emits a light for irradiating a portion of the pattern 200 including the fixed point Fx. The fixed point Fx is located at a position that is not influenced by a displacement of the leaf plate 141m, and has a certain area. A setting position and an emitting direction of the light emitting unit 211 are fixed so as to set up the fixed point Fx along an arc direction of the outer side of the leaf plate 141m and to include a part region of the pattern 200. An imaging unit 212 is provided so as to image a region including the fixed point Fx on the pattern 200. Thus, the imaging unit 212 acquires the images of fixed points at a prescribed interval by time sequentially receiving reflected lights from the fixed point Fx. In the images of fixed points, the pattern 200 provided on the leaf plate 141m partially exists. The moving amount calculation unit 213 specifies a position of a special pattern 200a included in the pattern 200 and calculates a moving amount of the leaf plate 141m based on time sequential displacements of the special pattern 200a acquired through the imaging unit 212. The moving amount calculation unit 213 analyses images of a plurality of fixed points acquired in a time sequence in accordance with movements of the leaf plate 141m and acquires a moving amount of the leaf plate 141m. Since the pattern 200 fixed on the leaf plate 141m in accompany with the movement of the leaf plate 141m, each position of the fixed point Fx on the pattern 200 relatively changes in the plurality of fixed point images. Thus, by judging the position displacements of the fixed point Fx on the plurality of fixed point images time sequentially acquired through the moving amount calculation unit 213, the leaf plate position detection unit 210 can detect a moving amount of the leaf plate 141m. Despite of displacement amount of the leaf plate 141m, a part of or the whole of the special pattern 200a is desired to exist in the image of the fixed point. Thus, the pattern 200 formed on the leaf plate 141m is desirable to be a small as possible. In the present embodiment, the special pattern 200a is formed by a square of 0.5 mm in four directions as one unit of the pattern 200 for making a smaller area than an imaging area of the imaging unit 212. According to the radiation therapy equipment in consistent with the present embodiment, since each leaf plate is constantly tensioned in both a closing direction and an opening direction, an irradiation field can be accurately set up with substantially avoiding backlashes due to the driving gears. According, unnecessary irradiation of radiation on normal tissues can be protected. Even when a wire for tensioning a leaf plate in an opening direction is cut, the leaf plate moves in a closing direction by the force of the constant force spring to protect an object from unnecessary irradiation. Thus, a safety of the patient can be constantly protected. In the present embodiment, since a plurality of constant force springs is fixed on a pair of rotation shafts for driving the same number of the constant force springs, it becomes possible to reduce the numbers of the shafts for rotating the leaf plate driving gears comparing to the conventional radiation therapy equipment. Further, since each wire connected to each leaf plate can change direction through pulleys, it becomes possible to freely place a plurality of roll-up motors at appropriate positions in a limited space. Thus, an increased number of roll-up motors can be installed in a limited space of the irradiation unit. Consequently, it becomes possible to increase the number of leaf plates can so as to set up the irradiation field in a higher accuracy. According to the present embodiment, since a displacement and a position of a leaf plate can be detected by a non-contact system, displacements or detection errors of plate positions due to backlashes and wearing of toothed wheels are eliminated. Consequently, each position of a leaf plate can be accurately detected, and each irradiation field can be set up in a high accuracy. In place of the motor used in the embodiment, it is also possible to roll up the wire connected to the leaf plate by using powers of hydraulic pressure or air pressure. While a position of the leaf plate is directly detected by the leaf plate position detection unit in the embodiment, it is also possible to detect the leaf plate position by a non-contact detection system. For instance, a multi-pole magnetic pattern is attached on a side surface of a driving gear. And a magnetic sensor provided near the driving gear detects the magnetic pattern. Based on detected data through the magnetic sensor, each moving amount of a leaf plate can be controlled. While certain embodiments have been described, these embodiments are presented by way of example only, and are not intended to limit the scope of the invention. Indeed, the novel instruments described herein may be embodied in a variety of other forms; furthermore, various omissions and changes in the form of the instruments described herein may be made without departing from the spirit of the inventions. The accompanying claims and their equivalents are intended to cover such forms or modifications as would fall within the scope and spirit of the inventions.
047568798	claims	1. A core cover plug of a liquid metal-cooled nuclear reactor having a system of vessels sealed by a slab provided with two rotary plugs one of which is smaller than the other, said core cover plug being suspended on the smaller rotary plug, wherein the core cover plug comprises a structure constituted in part by suspension elements fixed to a lower part of the smaller rotary plug, and in part by control rod sleeve elements, said elements being joined, at least at the horizontal level, by vertical metal plates forming a honeycomb grid, and a conical metal deflecting plate positioned directly above the core and fixed to certain of said sleeve elements by a radial clearance expansion system, and wherein the deflecting plate supports a group of tubes for sampling liquid metal at the outlet from the core in order to located possible sheath fractures and for locating thermocouples, said group being rigidified by two other cone-shaped plates which, with the deflecting plate, contribute to the deflection of the liquid metal jet leaving the core, wherein said cone-shaped plates are located below said vertical metal plates. 2. A core cover plug according to claim 1, wherein the suspension elements are control rod sleeves. 3. A core cover plug according to claim 1, wherein the suspension elements have a square cross section. 4. A core cover plug according to claim 1 wherein the suspension elements have a cross-shape cross section.
	summary
063209233	claims	1. A locking device/for use in a nuclear reactor to restrain a jet pump downcomer comprising: a restraining ring; a downcomer carried by the restraining ring; a wedge supported by the restraining ring and the downcomer; a guide bolt extending in an axial direction in relation to the downcomer such that the wedge is movably supported by the guide bolt; a locking device comprising: a restraining ring; a downcomer carried by the restraining ring; a wedge supported by the restraining ring and the downcomer; a guide bolt extending in an axial direction in relation to the downcomer such that the wedge is movably supported by the guide bolt; and a locking device comprising: wherein the means for inducing a compressive force against the upper jaw and the lower jaw is a tightening screw extending between the upper jaw and the lower jaw. 2. The locking device as defined in claim 1, wherein the wedge is interposed between the upper jaw and the lower jaw. 3. A locking device as defined in claim 1, wherein the guide bolt is carried by two guide blocks supported by the downcomer. 4. The locking device as defined in claim 1, wherein the guide blocks are welded to the downcomer. 5. The locking device as defined in claim 1, wherein one end of the upper jaw and a corresponding end of the lower jaw are moveably supported by the guide bolt. 6. The locking device as defined in claim 5, wherein the upper jaw includes a hook-shaped member at the distal end thereof. 7. The locking device as defined in claim 5, wherein the lower jaw includes a hook-shaped member at the distal end thereof. 8. The locking device as defined in claim 1, wherein the tightening screw is retained within a housing. 9. The locking device as defined in claim 8, wherein the housing defines a recess such that the lower surface of the recess is brought into contact with the restraining ring. 10. The locking device as defined in claim 1, wherein the locking device is fabricated of stainless steel. 11. A locking device for use in a nuclear reactor to restrain a jet pump downcomer comprising:
	summary
047755091	summary	BACKGROUND OF THE INVENTION Technical Field The invention relates generally to fuel assemblies for nuclear reactors of the type incorporating a bundle of fuel rods disposed at the nodal points of a regular array and spacer grids distributed along the bundle. The invention relates more particularly to grids for use in fuel assemblies, which include at least two series of mutually parallel plates or straps defining cells which receive fuel rods and possibly elements substituted for them at certain locations in the array. Although the invention is general in scope, it is particularly suitable for use in fuel assemblies whose rods are arranged in a triangular array whose pitch is "tight", i.e., only a little greater than the diameter of the rods. The use of such a triangular lattice is desirable in an undermoderated reactor core, which means that the coolant streams between the rods must be narrow. However, the construction should allow the required coolant flow and provide adequate mixing of the coolant streams, without causing an excessive head loss. The grids of nuclear fuel assemblies for water moderated and cooled reactors, and particularly PWRs, generally retain the rods at the nodal points of a square array. The grids typically include a girdle and two sets of orthogonal plates. A fuel assembly has also been proposed U.S. Pat. No. 3,068,163 having grids for maintaining the rods at the nodal points of a square array, which grids are formed by a flat or undulating endless strip passing between the rods, with intermeshing at the crossing points. This solution, if applied to a "tight" pitch fuel assembly, does not allow satisfactory coolant flow, conditions, mixing of the different streams and an acceptable pressure loss. Support grids for retaining a triangular array of fuel rods have also been proposed, for instance in European No. 0 065 613 (Downs), however including a single layer of intersecting corrugated strips arranged along two different directions. SUMMARY OF THE INVENTION It is an object of the invention to provide a grid for a nuclear fuel assembly which retains the rods efficiently and accurately and achieves mixing of the fluid streams, while including a small amount of neutron-absorbing material. To this end, there is provided a spacing grid for a nuclear fuel assembly including a peripheral girdle and at least two sets of mutually parallel plates, the plates of one set being oblique with respect to the plates of the other set so that the sets define cells for the rods passing therethrough, wherein the parallel plates are distributed in at least two beds spaced apart in the longitudinal direction of the assembly. In a fuel assembly having a triangular array of fuel rods, the plates of each set will be generally parallel to a respective one of the imaginary lines joining the centers of aligned rods, so that three sets of plates will be required. In a first embodiment, the grid supports the rods: the plates are then provided with rod support means, such as bosses, dimples or tongues, for supporting each rod in each bed at at least two diametrically opposed points, the support points of a rod in a grid being distributed between several beds in which the plates belong to different sets. In other words, the support points of the same rod on the plates of one bed are offset angularly with respect to the supports by the plates of another bed. In another embodiment, the grid is mainly intended to fulfill essentially a function of mixing the fluid streams. Then the two beds of the grids, (or two beds at least if the grid includes more of them) are provided with half-fins having different angular positions in the two beds. In such a grid, whose function is essentially thermo-hydraulic, the plates may be thin and have a small height, as compared with the height of the usual grids which support the rods, center them and have a mixing function. With the arrangement of the plates in several successive beds, the pressure loss is reduced as compared to that caused by a grid where all plates are in the same plane, for the same cumulative cross-section of all grids. Because the plates of several beds include mixing fins, the latter acquire a three-dimensional effect. Depending on the type of array and the arrangement of the fins, different types of mixing and flow distribution may be provided by selecting the most appropriate arrangement in each case. In a grid having a mixing function, the girdle may be formed of several successive sections each associated with a bed, which sections are joined together solely by longitudinal junction elements, such as rods placed in the angles of the girdle. In yet another embodiment, the grid simultaneously fulfills all different functions mentioned above. The plates may be formed of metal strips which are flat or, for increased strength, corrugated to the spacing pitch of the rods along the plates. The metal strip, whether flat or undulated, may include bosses directed from one side and/or from the other side of the plate, towards the centers of the cells defined by the plate. The invention also provides a fuel assembly of the above defined type in which some of the grids at least, intended to support and center the rods, are secured to tubes connected to end nozzles for forming an assembly skeleton. Some at least of the grids then have one of the above defined constructions providing support and are connected to the tubes of the skeleton by sleeves each of which is common to all beds in the grid.
039716970	claims	1. In combination with a high energy accelerator having a duct containing a beam of particles having high energy between 200 MeV and 2000 MeV and current sufficiently high that the total beam power is greater than two kilowatts, apparatus for producing .sup.123 I comprising a heat pipe comprising a tubular container extending into said duct, a portion of said tubular container being in the path of said high energy beam, a supply of cesium 133 in said portion of said tubular container in said beam, said cesium 133 being bombarded by said beam thereby vaporizing the same and producing radioactive isotopes of xenon including .sup.123 Xe and contaminants selected from the group consisting of iodine, tellurium, antimony, tin, indium and cesium by spallation, cooling means around a portion of said tubular container remote from said portion of said beam for condensing said vaporized cesium, a first trap connected to said heat pipe for receiving said radioactive isotopes of xenon and the contaminants after the same have vaporized and removing said radioactive isotopes of said contaminants, and a second trap connected to said first trap for receiving said radioactive isotopes of xenon therefrom and removing the same, said second trap forming a container for holding said radioactive isotopes of xenon for a period of time sufficient for the .sup.123 Xe to decay to .sup.123 I. a beam of protons having an energy greater than 200 MeV, a heat pipe having one end extending into said beam and the other end in communication with said cold traps, and a supply of cesium 133 target material in said one end of said heat pipe whereby said cesium 133 is bombarded by said beam thereby vaporizing the same and producing the radioactive isotopes of xenon and contaminants by spallation. 2. Apparatus as claimed in claim 1 wherein the first trap contains solid carbon dioxide. 3. Apparatus as claimed in claim 1 wherein the cooled portion of the heat pipe extends outside the duct. 4. Apparatus as claimed in claim 1 including a porous metal plug in said heat pipe adjacent to said cooling means for preventing cesium from being transported to said first trap. 5. Apparatus as claimed in claim 1 wherein the second trap contains a pharmaceutical compound whereby said pharmaceutical compound is tagged when the .sup.123 Xe decays to .sup.123 I. 6. In apparatus for producing .sup.123 I wherein radioactive isotopes of xenon including .sup.123 Xe and contaminants selected from the group consisting of iodine, tellurium, antimony, tin, indium and cesium are passed through cold traps to sequentially remove the radioactive isotopes of contaminants and xenon, the improvement comprising
	description	1. Field of the Invention The present invention relates generally to the field of nuclear reactors and in particular to a radioactive debris trap to be installed in the primary outlet plenum of a steam generator for removing fine particles and chips of metal from the primary heat transport system of a nuclear power plant. 2. Description of the Related Art Referring to FIG. 1, and as described in Steam/its generation and use, 40th Edition, Stultz and Kitto, Eds., Copyright©1992, The Babcock & Wilcox Company, and Steam/its generation and use, 41st Edition, Kitto and Stultz, Eds., Copyright©2005, The Babcock & Wilcox Company, recirculating steam generators (RSGs) used in nuclear power plants are supplied by a number of manufacturers worldwide as part of pressurized water reactor (PWR) or pressurized heavy water reactors (PHWR) (mainly CANDU) systems. These are large devices, ranging in height from approximately 38 to 73 ft (11.6 to 22.3 m) and weighing from approximately 50 to 790 tons (45 to 717 tm) each. Each RSG is a vertical shell, inverted U-tube heat exchanger with steam-water separation equipment located above the tube bundle inside the upper shell (or steam drum). A cylindrical shroud or bundle wrapper surrounds the tube bundle separating it from the lower shell. This creates an annular region which serves as the downcomer to return the recirculated water from the steam separators to the tube bundle inlet at the bottom of the unit. In a feed ring type RSG, generally designated 100 as illustrated in FIG. 1, feedwater is introduced by a nozzle and header to the top of the downcomer, and flows with the separator return flow down and into the tube bundle. In a preheater type RSG, feed flow enters the steam generator through a nozzle and feedwater distribution box to the baffled section at the cold leg outlet end of the tube bundle where it is heated to saturation before joining with the hot leg riser flow within the tube bundle. The flow configuration and the major design features of a typical feed ring type RSG are as follows. The hot primary coolant enters a portion of the vessel primary head 110, via primary inlet nozzle 120, which is separated into two plenums 130, 140 by a divider plate 50. The primary coolant flows through the inside of the U-tube bundle 150 and exits the steam generator 100 through the primary head outlet plenum 140 and primary outlet nozzle 160. In most RSG designs, the U-tubes make a continuous 180 degree bend at the top of the tube bundle. In the configuration shown, secondary-side feedwater enters the upper shell 170 via feedwater nozzle 180 and is conveyed to a feed ring (not shown) and is mixed with water returning from the steam-water separation equipment 190 located in the upper shell 170. The water flows down the downcomer annulus between the shroud and the shell to the tubesheet where it enters the tube bundle. The secondary-side water is heated as it passes up through the tube bundle generating steam through nucleate boiling heat transfer, creating a two-phase flow. Steam of 10 to 40% quality, depending on hot-side or cold-side U-tube bundle location, exits the tube bundle and is distributed to the primary and secondary steam separation equipment 190 in the upper shell 170 to send effectively moisture-free (<0.25% water) steam to the secondary-side power cycle via steam outlet nozzle 200. Water leaving the steam separators is recirculated down the annulus where it mixes with the feedwater before being returned to the bundle inlet for further steam generation. During operation, debris can sometimes begin to accumulate in the primary coolant loop or primary heat transport (PHT) system of such steam generators 100. Depending upon the source of the debris, the type of debris which can typically be found in a PHT system can measure 1 square mm or less, or the debris fragments can be as large as 2 mm wide by 4 mm long. Damage and defects caused by debris can cause a problems for nuclear power plants. Thus, it logically follows that various debris trapping devices have been developed in response to the industry wide problems caused by debris. For example, U.S. Pat. No. 4,684,496 to Wilson et al. (“Wilson”) describes a debris trap for a pressurized water nuclear reactor to be installed into the reactor vessel itself. The debris trap disclosed in Wilson is mounted within a bottom nozzle of a fuel assembly so as to capture and retain debris carried by coolant flowing from the lower core plate openings of a nuclear reactor to a fuel assembly and is made up of a plurality of straps aligned with one another in a crisscross arrangement. However, due to the large scope of the problems caused by debris in nuclear power plants, there remains a clear need for a simple debris trap which can remove a greater amount of debris and reduce the problems caused by debris. One aspect of the present invention is drawn to a radioactive debris trap that is capable of targeting and removing a higher than average concentration of debris particles, i.e., fine particles and chips of metal, from a nuclear power plant's primary heat transport (PHT) system without disrupting all of the primary flow the system. Another aspect of the present invention is drawn to a simple yet effective radioactive debris trap which can be installed in any type of steam generator. Accordingly, one aspect of the present invention is drawn to a radioactive debris trap configured to be installed in a steam generator for the purpose of removing debris which is entrained in a primary flow of a nuclear power plant's PHT system. The debris trap is made entirely of metal and includes an outer cylinder having an outer surface, a top end and a bottom end having a perimeter. The invention also includes a coaxial inner cylinder located within the outer cylinder. The inner cylinder includes a top end and a bottom end. Additionally, a top plate connects the top end of both the outer cylinder and the inner cylinder. A plurality of small holes is located at the top end of the outer cylinder. Liquid contained in the primary flow exits the debris trap through these holes. Also, the plurality of holes are sized to ensure that the gravitational force exerted on the debris entrained in the primary flow is larger than the upward drag force exerted on that debris by the flow itself. The debris trap of the present invention includes an annular cavity which is located between the outer cylinder and the inner cylinder. The annular cavity contains a settling chamber positioned below the plurality of small holes. A bottom plate encloses the radioactive debris trap and is connected to the perimeter of the bottom end of the outer cylinder. A gap is located between the bottom end of the inner cylinder and the bottom plate. Primary flow with debris entrained therein enters the settling chamber through this gap. The present invention also includes a means for fixedly connecting the radioactive debris trap to the steam generator which is fixedly attached to the outer surface of the outer cylinder on the back side of the debris trap. Additionally a means for removing the radioactive debris trap from the steam generator without exposing personnel to excessive radiation is provided on the front side of the debris trap. The preferred embodiment for this means includes three female support brackets which are triangularly arranged with respect to each other and which include an uppermost centrally positioned female support bracket. The means for removing the radioactive debris trap from the steam generator also includes a remote robotic arm which is configured with three complementary male support brackets which engage the female support brackets attached to the front of the debris trap. The various features of novelty which characterize the invention are pointed out with particularity in the claims annexed to and forming a part of this disclosure. For a better understanding of the invention, its operating advantages and specific benefits attained by its uses, reference is made to the accompanying drawings and descriptive matter in which a preferred embodiment of the invention is illustrated. Referring now to the drawings, wherein like reference numerals are used to refer to the same or functionally similar elements throughout the several drawings, FIGS. 2 and 3 show a preferred embodiment of the radioactive debris trap 10 of the present invention. The present invention can be installed in the primary head of a U-tube type recirculating steam generator (RSG) 100 for removing fine particles and chips of metal from a primary flow 46 of a nuclear power plant's primary heat transport (PHT) system. As shown in FIGS. 2 and 3, among its major components, the preferred embodiment of the radioactive debris trap 10 which has a front side and a back side includes an outer cylinder 12 and a coaxial inner cylinder 14 positioned within the outer cylinder 12. Both the outer cylinder 12 and the inner cylinder 14 have a top end as well as a bottom end and the outer cylinder 12 has an outer surface. Additionally, a conical top plate 16 connects the outer metallic cylinder 12 to the inner metallic cylinder 14. The top plate 16 functions to direct debris toward the trap 10 and to prevent any particles from exiting the top of the trap 10. Alternatively, the top plate 16 can be a planar surface which may be flat or inclined In addition, a bottom plate 18 is attached to the entire perimeter of the bottom end of the outer cylinder 12 and hence closes the entire bottom of the trap 10. Also a means for fixedly connecting the radioactive debris trap 10 to the steam generator 100, preferably to the divider plate 50, is fixedly attached to the outer surface of the outer cylinder 12 on the back of the debris trap 10. For example, FIG. 2 shows three female support brackets 20 which are fastened to the outer surface of the outer cylinder 12. These female support brackets 20 are also triangularly arranged with respect to each other. The three female support brackets 20 serve the purpose of facilitating attachment to the steam generator 100 divider plate 50 shown in FIG. 1. However, other means of attachment may be used. For example such means could include male support brackets 32, similar to those shown in FIG. 5 and discussed in greater detail below, which are attached to the debris trap 10 and which are designed to engage with female supports. Also a design which uses bolted connections is another potential means of attaching the debris trap 10 to the steam generator 100. As shown, a similar set of support brackets 20 are provided on an opposite side of the debris trap 10 to facilitate its installation and removal using either manual or robotically controlled tooling, as described later in this specification. A plurality of small holes 24 which are sized to ensure low upward velocity in the settling chamber are located at the top end of the outer cylinder 12. Liquid contained in the flow 46 exits the debris trap 10 through these holes 24, while debris swept into the debris trap 10 along with the primary flow 46 remains trapped therein. The preferred embodiment of the radioactive debris trap 10 includes a gap 22a which is located between the bottom of the inner cylinder 14 and the bottom plate 18. This gap 22a allows the flow 46 to sweep the debris particles outward into a settling chamber 48 which is located in an annular cavity positioned between the outer cylinder 14 and the inner cylinder 12. As explained in more detail below, upward fluid velocity in the settling chamber is kept below the settling velocity to cause the debris that enters the debris trap 10 to settle out and remain in the settling chamber 48 or on the bottom plate 18. The preferred embodiment of the debris trap 10 is made entirely of steel of a grade suitable to withstand the temperatures, stresses, flow conditions and chemistry conditions encountered in the steam generator. FIGS. 4 and 4A illustrate another embodiment of the debris trap 10 according to the present invention. The embodiment shown in FIGS. 4 and 4A includes an open-topped box 36 having a bottom 38, a front side 40, a back side 42, a right side and a left side. The front side 40 of the open-topped box 36 has a top end and a bottom end and a back side 42 of the open-topped box 36 has a top end, a bottom end as well as an outer surface. Although FIG. 4 shows the open-topped box 36 having a rectangular shape, the debris trap 10 can take any shape which will not prevent it from accomplishing its desired function. For example, the open-topped box 36 can be square. Moreover, the embodiment of FIGS. 4 and 4A also includes an angled plate 44 located inside the open-topped box 36. The angled plate 44 is positioned to direct the flow 46 and the debris entrained therein into the debris trap 10 and it has a front end, a right side, a left side and a back end. The front end of the angled plate is fixedly attached to the front side 40 of the open-topped box 36 and is flush with the top end of the box 36. The right and left sides of the angled plate 44 are fixedly attached to the right side of the open-topped box 36 and the left side of the open-topped box 36 respectively. The back end of the angled plate 44 is proximate yet unconnected to both the bottom of the open-topped box 36 and the back side 42 of the open-topped box 36. Between the back end of the angled plate 44 and the back side 42 of the open-topped box 36 is provided a flow slot 22b. It is at this point that the flow 46 enters the debris trap 10. Additionally similar to the preferred embodiment, the embodiment shown in FIGS. 4 and 4A includes a means for fixedly attaching the radioactive debris trap 10 to the steam generator divider plate 50 attached to the outer surface of the back side 42 of the open-topped box 36, as well as means for facilitating its installation and removal using either manual or robotically controlled tooling, as described later in this specification. FIG. 4 shows the three female support brackets 20 fastened to the back surface 42 of the debris trap 10. Furthermore, the debris trap 10 of the present invention is designed to employ the same arrangement of female support brackets 20 on the front as well as the rear of the trap 10 to securely engage the male support brackets 32 on both the debris trap support 26 for attachment as well as on a robotic arm R removal tool, discussed below. Additionally, more or fewer support brackets 20 could be used. Alternatively, male support brackets 32 could be attached to the debris trap 10 which are designed to engage with female support brackets 20. Also designs using bolted connections are another potential means of attaching the trap 10 to the steam generator. Additionally as illustrated in FIGS. 7A and 7B, both of the above mentioned embodiments of the present invention may include a bolt 70 and a locking tab 74 for engaging the bolt hex head 72 to ensure the trap 10 cannot move upwards and become disengaged from the steam generator divider plate 50. Under certain circumstances, the bolt 70 may not be required if analysis and flow conditions indicate that upward movement of the debris trap 10 is unlikely. The locking tab 74 is fabricated from sheet metal. The uppermost center bracket of both the male brackets 32 and the female brackets 20 are fitted with a pilot hole 60 to engage the bolt 70. Additionally, the locking tab 74 includes a lip to fold over the edge of the male or female bracket 32, 20 so the bolt 70 cannot turn. Alternatively, the bolt 70 can also be welded to the brackets 20, 32 or to a plate so it will not turn and then bent over the hex head portion 72 of the bolt 70 to prevent the bolt 70 from turning. Also as seen in FIGS. 6 and 6A, an angle is machined on the face of the support brackets 20, 32 to ensure a snug fit. Additionally, the embodiment of the present invention shown in FIG. 4 also includes a plurality of small holes 24 positioned across the top of the front face 40 of the open top box 36. These small holes 24 are positioned on top of the settling chamber 48 and as explained in more detail below, they are sized to ensure low velocities in the settling chamber 48. All embodiments of the debris trap 10 of the present invention are securely fixed to a support bracket that is welded or bolted to an existing divider plate 50 in a steam generator 100 as shown in FIG. 1. The divider plate 50 is attached to the side of the primary head and to the bottom of the tubesheet and directs the flow 46 from the primary inlet nozzle 120 through the U-tubes 150 in the steam generator 100 back into the primary outlet plenum 140 and into the primary outlet nozzle 160. Moreover, for steam generators 100 with bolted type divider plates (not shown), the support bracket can be designed to use existing bolting arrangement holes. Moreover, the inventive debris trap 10 is securely fixed to a means for connecting the debris trap 10 to the divider plate 50 or at another location in the primary outlet plenum 140 of the steam generator 100. In the preferred embodiment, this means is attached to the outer surface of the outer cylinder 12. In the embodiment shown in FIGS. 4 and 4A, this means is attached to the outer surface of the back side 42 of the open-topped box 36. As shown in FIG. 5 the means for attaching the trap 10 to a steam generator may take the form of a debris trap support base plate 28 which has a back surface that is fixedly attached to a divider plate 50 of the steam generator 100. The debris trap support base plate 28 as shown in FIG. 5 also includes three male support brackets 32 which supports the debris trap 10 and are designed to engage with the female support brackets 20 shown in FIGS. 2, 3, 4 and 4A. As illustrated in FIGS. 6 and 6A, to ensure the trap 10 is securely fastened to the debris trap support base plate 28 the preferred embodiment includes angled faces on the male and female support brackets 32, 20 to ensure a snug fit. As described above, fasteners may be required to secure the debris trap 10 to the debris trap support base plate 28. The preferred embodiment for the fastener is a bolt with a thin sheet metal locking tab, which is required to ensure the bolt does not come loose, that is screwed to a threaded connection on the upper male support bracket and hence ensures the trap cannot move upwards and become disengaged from the supports. Any reasonable means of attachment which is sufficiently strong can be used, if required. Also, shown in FIG. 5 are four pilot holes 34 which could accommodate fasteners such as bolts or rivets. The preferred embodiment would be bolts with locking tabs as this design has proven operating experience. Alternatively, the trap 10 can be welded to the divider plate 50, but this would make removal more difficult. It will be appreciated that while FIG. 5 illustrates a debris trap support base plate 28 suitable for the FIGS. 4 and 4A embodiment, support brackets on the debris trap support base plate 28 suitable for the FIG. 3 embodiment would be angled or curved as necessary to accommodate the curved outer surface of the debris trap 10. As mentioned above, both embodiments of the inventive debris trap 10 discussed include three female support brackets 20b attached to their front sides to facilitate removal using remote tooling, such as a long rod or a robotic arm R fitted with male brackets 32 (see FIG. 8) which are complementary to the female support brackets 20b and which are designed to engage the female support brackets 20b. However, as with the preferred embodiment, other means of attachment may be used. Additional remote tooling consisting of a long rod or a robotic arm with an impacting hammer that can apply a vibratory force to the bottom of the debris trap 10 may be required to loosen the connection between the male and female support brackets 32, 20. Furthermore, other suitable devices which can safely remove the debris trap 10 without exposing personnel to excessive radiation may be used. For example a robotic arm R with a clamping device could be used that would apply a force to the side of the debris trap 10 could be used. If the radiation field is sufficiently low personnel with protective clothing could be used to remove the debris trap. The operating principle for the debris trap 10 is to allow the flow 46 containing the debris to enter the top of the debris trap 10 and enter the settling chamber 48 through the gap 22a or the slot 22b. After fluid enters the trap 10 it changes direction before it flows slowly upwards and exists through the plurality of small holes 24 near the top of the trap 10. Provided that the vertical velocity of the flow 46 in the trap 10 after the flow 46 turns upwards is less than the settling velocity the particle will stay in the trap 10. In other words the condition which causes retention of the particles within the trap 10 is the downward force on the particle due to gravity being larger than the upward drag force on the particle exerted by the flow 46 of primary fluid within the debris trap 10. The pressure differential available to cause flow 46 in the debris trap 10 is a function of fluid impingement velocity and the lower static pressure at the exit of the plurality of small holes 24. Thus, the multiple small holes located near the top of the settling chamber 48 are sized to control the upward velocity in the settling chamber to less than the particle settling velocity. That is to say that they are sized to ensure that the gravitational force exerted on the debris entrained in the primary flow 46 is larger than the upward drag force exerted on the debris by the flow 46 itself. This ensures that debris which enters the trap will settle to the bottom of the trap. The settling chamber is the portion of the debris trap upstream of the vent holes. Besides capturing radioactive debris other key design requirements for the debris trap 10 of the present invention are as follows. Firstly it must stay securely in place during operation. In addition, it must not have any parts that can become loose within the PHT system. It is also essential that the debris trap 10 fit through the manway in the primary head of a steam generator for easy installation. The inventive debris trap 10 must also be light, preferable less than 20 lb (9 kg) and hence easy for one person to install. Furthermore, it must be easy to enclose in a dust tight radioactively shielded envelope to allow handling and removal for safe storage. It must also be easy to remove and easy to transport to the storage site for radioactive waste after the radioactive debris has been captured. Finally it must be small enough so it does not significantly increase the primary heat transport system flow losses. The inventive radioactive debris trap 10 is suitable for installation in any steam generator and is capable of removing radioactive debris that may be present in any nuclear power plant PHT system. The radioactive debris trap 10 could be installed during a typical boiler outage, while the head is open and PHT system is already in a low-level drain state. The outlet side is chosen not only because it is less turbulent than the inlet, but also because the flow from a few select tubes that contain a higher than average concentration of debris particles may be targeted to be “partially filtered” without disrupting all of the primary flow. The key advantage of installing a debris trap inside the boiler is that the device has a very high probability of capturing most of the particles over the period of one reactor operating cycle of say 8000 hours. It is expected that the debris will be more concentrated in areas of direct primary inlet flow impingement, including tube ends near the center of the steam generator and near the divider plate 50. By locating the debris trap below the outlet ends of such tubes it is expected that the flow entering the debris trap will contain a relatively concentrated stream of debris which is available for capture within the debris trap. During development of the inventive trap 10 tests were conducted by dropping small pieces of wire which was 0.017″ diameter by 0.017″ long into a container filled with 55° F. water. A stop watch was used to time how long it took the particles to drop to the bottom of the container. The average free fall velocity during the test was 0.61 ft/sec. The small holes at the outlet of the debris trap are sized to achieve a vertical velocity in the debris trap of 0.3 ft/sec which is half the settling velocity in the test. Larger particles tend to have a higher settling velocity than small ones. Hence this test confirms the debris trap will capture all particles greater than 0.017″ in diameter, which is the majority of the particles large enough to cause problems. Additionally, an approximate efficiency analysis may be performed using as an example, a power plant having eight steam generators per reactor. Each steam generator has 4200 tubes for a total of 33,600 tubes in all the steam generators. There are two reactor coolant pumps for each bank of four steam generators. Assuming the radioactive debris trap 10 of the present invention is installed in two steam generators, (one steam generator in each group of steam generators supplied by the same pumps), it is expected that over time most of the debris will be removed. The debris trap will filter a small fraction of the fluid coming from the ends of the tubes directly above the debris trap. Table 1 shows the expected particle removal rate based on the assumption that debris is uniformly distributed within the primary fluid. TABLE 1Particle Removal with Debris Trap InstalledParticles in PHT% of particlesTime (hours)system (kg)removed03 0%502.0831.3%1001.4352.8%2000.6777.7%4000.1595.0%8000.0199.8% A Computational Fluid Dynamic (CFD) analysis that includes particle tracking throughout the system will be used to more accurately predict the distribution of particles in the boiler tubes and in the debris trap as a function of time. While specific embodiments of the present invention have been shown and described in detail to illustrate the application and principles of the invention, it will be understood that it is not intended that the present invention be limited thereto and that the invention may be embodied otherwise without departing from such principles. In some embodiments of the invention, certain features of the invention may sometimes be used to advantage without a corresponding use of the other features. Accordingly, all such changes and embodiments properly fall within the scope of the following claims.
	summary
042726824	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to an ion milling machine and more particularly to an ion milling machine specimen elevator and airlock mechanism which facilitates specimen viewing and specimen exchange. 2. Description of the Prior Art Ion milling machines are used in the preparation of electron microscope specimens. An article entitled "An Improved Ion Thinning Apparatus" by A. H. Heuer, et al. from The Review of Scientific Instruments, Vol. 42, No. 8, Aug. 1971, pp. 1177-1184, which is herein incorporated by reference, provides good background information on ion milling. Ion milling machines are employed for thinning specimens to a thickness of the order of 0.5 micrometers for examination by transmission electron microscopy. Desired thinning is accomplished by placing the specimen in the path of one or more beams of energetic ions and neutral atoms which sputter atoms from the specimen surface. Although material removal is very slow during ion milling, this procedure does much less damage to the underlying specimen than other material removal methods such as cutting or grinding. Ion milling has become widely used as a method for electron microscope specimen preparation, especially for those materials that because of their chemical nature cannot be thinned by electro-polishing or chemical polishing methods. Ion thinning is accomplished in an evacuated chamber; usually the specimen is rotated during exposure to the ion beams to improve the uniformity of thinning across the specimen surface. In order to study the progress of the ion milling operation, the specimen must be periodically examined using a lower power microscope. This inspection is difficult in prior art ion thinning machines because material sputtered from the specimen and specimen holders coats the observation windows and sources of illumination, thereby obscuring the view of the specimen. Further, because the specimen is positioned near the center of the work chamber and is surrounded by the ion guns and other related mechanisms, the specimen must be viewed at an uncomfortably long distance. A second deficiency of presently available ion thinning units is that the entire work chamber must be raised to atmospheric pressure before specimens can be exchanged. Raising the work chamber to atmospheric pressure has at least three disadvantages: (1) the time for specimen exchange is lengthy because a large volume of air has to be evacuated before the ion guns can be turned on; (2) the ion guns are initially less stable after exposure to atmospheric pressure and must be pumped for a long period of time before they restabilize; and (3) a costly valving mechanism is needed to isolate the work chamber from the pumping system. In presently available ion thinning units, it is awkward to load and unload specimens. This is disadvantageous because electron microscope specimens are extremely fragile and are easily damaged by the small mechanical shocks they receive during the loading and unloading operation. SUMMARY OF THE INVENTION The present invention provides a specimen elevator and airlock apparatus which is quick, gentle and convenient in operation and solves many of the problems associated with prior art ion milling apparatus. The specimen elevator for the disclosed ion milling machine comprises a vertically elongated piston which supports a specimen holder at its top. The elongated piston is movable between a raised and lowered position. When the elongated piston is in the lowered position, it is rotated about its vertical axis. The elongated piston extends through an O-ring seal into an evacuated work chamber where the ion milling is accomplished. The piston can be moved either up or down by altering the pressure in a pneumatic cylinder mounted at the bottom of the work chamber around the piston. A sealing plate is disposed around the piston and provides for sealing between the piston and inner diameter of the cylinder walls and allows the piston to be rotated around its vertical axis. A motor is mounted at the bottom of the cylinder for rotating the piston when it is in its lowered position. During ion milling the piston is at its lowered position and is held there by pressurizing the pneumatic cylinder. With the piston in its lowered operating position, the specimen lies inside the evacuated work chamber. If it is desired to remove the specimen from the work chamber without disturbing the work chamber vacuum, the pressure in the pneumatic cylinder is released and the piston automatically moves to its raised position under the force exerted by atmospheric pressure. As the piston is forced upwardly by atmospheric pressure, it passes through a second O-ring pressure seal into a small upper chamber. As the piston moves the specimen holder into the small upper chamber, a seal is provided between the upper chamber and the work chamber. Thus any air admitted to the small upper chamber will not pass to the work chamber. Means are provided for admitting atmospheric pressure to the small upper chamber. The small upper chamber can be lifted from the top of the large work chamber when the internal pressure in the small chamber rises to atmospheric. This exposes the specimen holder and thus permits removal of the specimen or close examination of the specimen to study the progress of the ion milling operation. The specimen is reintroduced into the work chamber by first replacing and evacuating the small chamber, and then repressurizing the bottom control cylinder. When the control cylinder is pressurized, the piston is forced to move downwardly to its lowered operating position. As the piston moves downwardly, the seal between the upper chamber and the working chamber is eliminated and the two chambers are in open communication. The small upper chamber is held to the working chamber by atmospheric pressure. A driven bevel gear is provided at the bottom of the control cylinder. A mating bevel gear is attached to the piston. When the piston is in its lowered position, its attached bevel gear engages the driven bevel gear causing the piston to slowly rotate. The specimen holder is mounted on the top of the piston and, therefore, specimens can be held in place simply by the force of gravity. This arrangement makes unnecessary the use of mechanical clamps and the consequent risk of physical damage to the fragile specimen. It is an object of the present invention to provide a mechanism which enables a specimen undergoing ion thinning to be raised quickly from the center of the work chamber to a more convenient viewing position and also to quickly lower the specimen back to its normal working position after inspection. It is another object of the invention to provide an ion milling machine having a specimen elevator device which raises the specimen to a viewing or exchange position and at the same time maintaining the working chamber in a sealed condition. It is yet a further object of this invention to provide a specimen handling apparatus for an ion milling machine which simplifies specimen handling and is quick and gentle in operation.
	summary
	abstract	Certain exemplary embodiments can provide a first isogrid defining a first plurality of zones, each zone from said first plurality of zones comprising a plurality of ligaments, each zone from said first plurality of zones defining a plurality of spaces, each space bounded by a first sub-plurality of ligaments from said plurality of ligaments, each of said ligaments comprising a plurality of ligament surfaces.
	summary
061817635	description	BEST MODE FOR CARRYING OUT THE INVENTION As illustrated in FIG. 1, a fuel assembly 10 comprises a plurality of laterally spaced fuel rods 12 supported between an upper tie plate 14 and a lower tie plate 16. 8.times.8, 9.times.9 or 10.times.10 arrays of fuel rods are typical but, for clarity of illustration, only some of the fuel rods 12 are shown. The fuel rods 12 pass through a plurality of vertically spaced fuel rod spacers 18 (one shown) which provide intermediate support to retain the elongated fuel rods 12 in spaced relation and to restrain the lateral vibration. Each of the fuel rods 12 is formed of an elongated tube containing a column of nuclear fuel 20. A plenum at the upper end of the fuel rod contains a spring 22 which maintains the column of fuel in position. The fuel rods 12 are sealed by upper and lower end plugs 24 and 26. The lower end plugs 26 are formed with a taper for registration and support within cavities 28 in the lower tie plate 16. Upper end plugs 24 are formed with extensions 30, the upper ends of which fit into support cavities in the upper tie plate 14. Several of the support cavities 28 in the lower tie plate 16 are formed with threads to receive the end plugs of certain fueled tie rods 12' having threaded end plug shanks 32. Extensions 34 of the end plugs 24 of these same fueled tie rods are elongated to pass through the cavities in the upper tie plate 14 are formed with threads to receive retaining nuts 36. Fitted on the extensions 34 between the upper end plugs 24 and the upper tie plate 14 are expansion springs 38. In this manner, the upper and lower tie plates in the fuel rods are formed into a unitary structure. The fuel assembly 10 further includes an open-ended, thin walled, tubular flow channel 40, of substantially square cross section, sized to form a sliding fit over the peripheral surfaces of the upper and lower tie plates 14, 16 and the spacers 18 so that the channel 40 can be mounted and removed from the fuel bundle without difficulty. Fixed to the top corner of the channel 40 is a tab 42 by which the channel 40 is fastened to a standard 44 of the upper tie plate 14 by screw 46. Since the channel 40 is not fastened to the lower tie plate 24, the upper end of the channel 40 is free to move with respect to the lower tie plate 16, in the event of movement of the upper end of the fuel assembly 10. The lower tie plate 16 is formed with a downwardly extending nose piece 48 which is tapered to engage a fuel assembly support socket (not shown). The lower end of the nose piece 48 is formed with an opening 50 to receive pressurized water so that it flows upward among the fuel rods. To aid in equalizing neutron moderation, the fuel assembly 10 is fitted with at least one large water tube 52 for conveying relatively cool water upwardly through the central region of the fuel assembly. The water rod 52, like the fuel rods, extends between and is supported by the upper and lower tie plates 14, 16, respectively. In this prior art arrangement, the water tube is provided with a plurality of holes 54 at its lower end which provide an inlet for water into the tube, while the upper end of the water tube is provided with a plurality of holes 56, 58 which provide an exit for the water flowing therein near the upper end of the fuel column 20 within the fuel rods. With this background, the discussion below with respect to the schematic drawings shown in FIGS. 2-8, all of which relate to water rod configurations, will be readily understood by those skilled in the art. The schematic diagram in FIG. 2 represents a water rod similar to that illustrated in FIG. 1. More specifically, the water rod 60 has a narrowed lower end 62 with one or more inlet holes 64, while the upper end of the water rod also includes a narrowed portion 66 with one or more outlet holes 68. The placement of these holes at the top and bottom of the water rod imposes the full bundle pressure drop to drive flow through the water rod. When reactor flow reduces, the pressure difference driving liquid through the water rod is also reduced. This configuration, however, maintains very little vapor formation even for low flow conditions. The water rod configuration in FIG. 3 has been developed to have varying amounts of steam in the water rod at different reactor flow conditions. More specifically, the water rod 70 has a narrowed lower end portion 72 with one or more inlet holes or apertures 74. At the upper end of the water rod, however, there is added a small diameter downflow extension tube 76 offset from the uppermost end of the water rod 70 by a horizontal extension 71 and extending downwardly to a location proximate the narrowed lower portion 72 of the water rod (and hence at the bottom of the fuel column). One or more outlet apertures or openings 78 are provided at the lower end of the extension rod 76. In this arrangement, water flows upward through a large path and then downward through the small extension tube for essentially the full length of the fuel in the adjacent fuel rods before reaching the one or more exit holes 78. Note that in this configuration, there is only a short axial distance between the inlet holes 74 and the outlet holes 78, with resultant small imposed pressure differential across these holes. For low flow conditions, the downward flow tube 76 is predominantly filled with steam, and the fluid in the upward path is supported like a standpipe with a low pressure differential. The resultant liquid content in the water rod 70 is thus quite low, being proportional to the imposed pressure differential. For normal operation, the small downflow tube 76 and significant outlet flow restriction combine to severely limit water rod flow. Thus, this design results in significant steam formation in the water rod with associated unfavorable fuel efficiency, under normal operation conditions. Turning now to FIG. 4, there is illustrated another recent water rod design wherein the water rod 80 has a narrowed lower end 82 with one or more inlet holes 84 and a narrowed upper end portion 86 with one or more outlet holes 88. In this configuration, however, a central standpipe 90 extends from the inlet openings 84 upwardly to a location proximate the narrowed upper end portion 86. With flow restrictions typical of current designs, sufficient water rod flow is allowed at normal operating conditions to avoid steam formation. For low flow conditions, it was contemplated that the annular region outside of the standpipe 90 would fill with steam when the imposed pressure differential drops below that necessary to spill liquid over the top of the standpipe. Analyses have indicated, however, that under such conditions, liquid will flow backward through the upper outlet hole or holes 88 and refill the annular region 92 outside the standpipe 90. Since this region 92 has no bottom drain, it can potentially collect even more liquid than current water rod designs under similar conditions. The configurations illustrated in FIGS. 3 and 4 highlight the difficulty in designing water rods that achieve negligible vapor content at normal reactor operating conditions, while providing sufficient vapor content at low reactor flow rates. In connection with the present invention, it has been determined that the locations of the inlet and outlet holes, as well as the flow areas and hydraulic characteristics of the upflow and downflow paths are most important. The imposed pressure differential across an SWR is in fact determined by the placement of the inlet and outlet holes relative to the fuel bundle. Designs with higher imposed pressure differential will cause the SWR transitions to occur at lower reactor flow rates. One method for changing the imposed pressure differential is by locating the inlet holes above or below the lower tie plate (UTP). The latter configuration adds the LTP pressure drop to the imposed pressure differential on the SWR. Another method for changing the imposed pressure differential is by varying the elevation of the outlet holes. The imposed pressure differential increases as the outlet holes are moved further up the fuel bundle. However, since this also shortens the length of the downflow region (which is steam filled during standpipe mode of operation) it results in somewhat higher liquid content in the SWR during standpipe mode of operation. Ultimately, raising the outlet hole elevation sufficiently will result in SWR designs that have no significant improvement over current designs. Conversely, lowering the outlet holes too near to the inlet can result in unfavorable designs that are unable to transition back from standpipe mode to siphon mode, even at full reactor flow rates. Preferred locations for the downflow outlet holes fall in the range 35% to 65% of the fuel column height. Within that range, however, outlet holes should be located just above the fuel bundle spacers (element 18 in FIG. 1). Since pressure changes between spacers are relatively small compared to local spacer losses, placing outlet holes just above spacers provides added downflow length for a small penalty in imposed pressure differential. Turning now to FIGS. 5-8, specific exemplary siphon water rods in accordance with this invention are illustrated. In the first exemplary embodiment shown in FIG. 5, the water rod 94 has a narrowed lower end portion 96 with one or more inlet apertures or openings 98 which are located adjacent and above the lower tie plate. At the upper end of the water rod 94, a return or downward flow tube 100 extends downwardly from the uppermost end of the water rod, similar to the extension 76 shown in FIG. 3. In this arrangement, however, the downward extension 100 is of significantly larger diameter and also terminates approximately midway along the length of the water rod 94 (and approximately midway along the length of the fuel columns in the fuel rods) with flow exiting one or more holes or exit openings 102. Raising outlet hole elevation increases SWR liquid content somewhat during standpipe mode and causes operating mode transitions to occur at lower reactor flows. As a practical matter, the return path or extension 100 could be contained within the cross sectional area of a single water rod. Such an arrangement is shown, for example, in FIG. 6 where the downward return tube (or second tubular member) 104 lies within the water rod 106 (or first tubular member); and with one or more outlet openings or apertures 108 located approximately midway along the water rod 106. The first tubular member 106 has a closed upper end. The extension 104 has an open top portion 110 so that fluid flowing upwardly through the water rod 106 can spill into the downward extension 104 and exit through the one or more apertures or openings 108. As in the previously described embodiments, the water rod 106 has a narrowed lower end portion 112 with one or more inlet holes or openings 114. Turning now to FIG. 7, another siphon water rod similar to that shown in FIGS. 5 and 6 is illustrated but wherein the return tube takes the form of an outer annulus. More specifically, the siphon water rod 116 of FIG. 7 includes a narrow lower end portion 118 with one or more inlet openings 120. Internally of the water rod 116, there is a narrowed upper end portion 122 which terminates at an open upper end 124. Surrounding the narrowed upper portion 122, is a substantially closed annular region 126 with one or more outlet exits or apertures 128 located substantially midway along the length of the water rod 116, just above a radial shoulder 130 where the narrowed upper portion 122 commences. FIG. 8 illustrates a siphon water rod configuration similar to that shown in FIG. 5. The shortened downflow tube of this invention permits usage in a fuel bundle assembly which incorporates part length fuel rods (PLR's) in the region below the downflow tube. Thus, the arrangement in FIG. 8 includes a water rod 132 with a narrowed lower end portion 134 with one or more inlet holes or openings 136. The downflow path is formed by a substantially similar diameter extension tube 138 having one or more exit openings or holes 140 at the lower end of the downflow path. Note that the exit hole or openings 140 are again located substantially midway along the length of the main water rod 132 (and approximately half way along the fuel columns within the fuel rods). In this arrangement, the fuel assembly includes one or more conventional partial length fuel rods 142 which terminate at a location proximate the outlet openings or holes 140. As indicated, the disclosed siphon water rods in accordance with this invention have bimodal states. The water rod operates in "siphon mode" when reactor flows are high enough for the imposed pressure differential to maintain the water rod filled with liquid. The water rod operates in "standpipe mode" when the imposed pressure differential decreases sufficiently to allow steam generation in the water rod to break the siphon effect. When the imposed pressure differential cycles from high to low and back to high (i.e., as reactor flow cycles down and up again), there is some hysteresis in the transitions between these bimodal states. Changing from the standpipe mode requires the imposed pressure differential to be greater than the density head of the upward path completely filled with liquid. Beyond that transition point, the siphon effect will cause the water rod flow to increase rapidly. However, once the siphon effect is operative, the siphon mode can be maintained even though the imposed differential is decreased somewhat below the prior transition point. While the invention has been described in connection with what is presently considered to be the most practical and preferred embodiment, it is to be understood that the invention is not to be limited to the disclosed embodiment, but on the contrary, is intended to cover various modifications and equivalent arrangements included within the spirit and scope of the appended claims.
	claims	1. A method for checking for leakage from tubular batteries, comprising:feeding tubular batteries with respective axial centers thereof aligned in parallel to each other to pass through a leakage check section placed opposite to a detection window of a leakage check mechanism;irradiating a sealed end face of the tubular battery in the leakage check section with an X-ray through the detection window and allowing a fluorescent X-ray coming out of the sealed end face to enter a fluorescent X-ray detector through the detection window; andanalyzing whether a fluorescent X-ray associated with an electrolyte component is contained in the incident fluorescent X-ray to thereby determine whether leakage occurs from the tubular battery, whereinthe detection window is defined in such a shape that a length thereof in a direction of feed of the tubular batteries is less than a spacing between the tubular batteries being fed, and a length thereof in an orientation orthogonal to the direction of feed is slightly larger than an outer size of the cross-sectional shape of the tubular batteries in an orientation orthogonal to their axial center. 2. The method for checking for leakage from tubular batteries according to claim 1, wherein occurrence of leakage is detected in accordance with an intensity per unit time of the fluorescent X-ray successively entering the fluorescent X-ray detector from each tubular battery which sequentially comes to oppose the detection window while being fed, or in accordance with an intensity per unit area of the sealed end face of the tubular battery. 3. The method for checking for leakage from tubular batteries according to claim 1, wherein: the detection window of the leakage check mechanism is disposed to oppose the sealed end face of the tubular battery being fed at a predetermined distance therebetween; a housing of the check mechanism contains an X-ray source for emitting an X-ray to a tubular battery, a mask for condensing a fluorescent X-ray emitted from the X-ray source into a beam, and the fluorescent X-ray detector upon which the fluorescent X-ray is incident; and the inside of the housing is kept in a helium gas atmosphere. 4. The method for checking for leakage from tubular batteries according to claim 1, wherein the tubular batteries are fed while being held on transfer disks in parallel to each other at regular intervals. 5. The method for checking for leakage from tubular batteries according to claim 4, wherein a housing which accommodates an X-ray source, a mask, and the fluorescent X-ray detector is installed in front of an apparatus casing, to which the transfer disks for feeding the tubular batteries are attached, so that the detection window provided on the housing opposes the transfer disks. 6. The method for checking for leakage from tubular batteries according to claim 5, wherein each of the tubular batteries is held in position on the transfer disks to pass through the leakage check mechanism so that a defective tubular battery which is determined to be leaky as a result of a check in the leakage check mechanism is rejected from the transfer disks onto a detectives collection path in order to be separated from a good-battery feed path. 7. The method for checking for leakage from tubular batteries according to claim 1, wherein an alkaline battery made up of an electrolyte containing a potassium hydroxide solution is checked to determine occurrence of leakage based on whether a fluorescent X-ray associated with a potassium component is contained in the fluorescent X-ray incident upon the fluorescent X-ray detector. 8. The method for checking for leakage from tubular batteries according to claim 3, wherein the mask formed of a metal that does not transmit an X-ray is allowed to condense an X-ray emitted from the X-ray source into a beam, and the beam is then transmitted through the detection window on the housing to the sealed end face of the tubular batteries being fed, and wherein at least a length of the detection window in the direction of feed of the tubular batteries is made variable.
	claims	1. A connector for fastening a steam generator to a flow mixing header, the connector disposed between a steam generator and a flow mixing header and fastening the steam generator to the flow mixing header in a sealing manner, the connector comprising:a base plate mounted on the flow mixing header and having a through hole formed at the center thereof;a steam generator connecting portion protruding along the circumference of the through hole in the base plate and allowing an outlet of the steam generator to be inserted and fastened thereto; anda plurality of elastic members provided in plurality along the circumference of the steam generator connecting portion on an other surface of the base plate and pressed toward a lower surface of the steam generator to be elastically deformed. 2. The connector of claim 1, further comprising:a horizontality adjusting portion provided in plurality along the circumference of the steam generator connecting portion on the other surface of the base plate to support a lower surface of the steam generator, and configured to adjust horizontality of the steam generator. 3. The connector of claim 2, whereinthe horizontality adjusting portion includes:a support protrusion outwardly protruding from the base plate and threaded on an outer circumferential surface thereof; anda height adjusting member fastened to the support protrusion by a screw and moving in a vertical direction by a rotational movement to adjust a height of the steam generator. 4. The connector of claim 1, further comprising:an escape preventing portion provided in plurality along the circumference of the steam generator connecting portion on the other surface of the base plate and connected to the lower surface of the steam generator so as not to escape from the steam generator. 5. The connector of claim 4, whereinthe escape preventing portion is connected to the steam generator to have a predetermined clearance. 6. The connector of claim 5, whereinthe escape preventing portion includes: a support member provided as a pair, protruding outwards from the base plate, and allowing a fixing protrusion formed in the steam generator to be inserted therebetween; anda fastening member inserted into the pair of support members and fastening the fixing protrusion to the pair of support members,wherein the fixing protrusion has a slit formed in a vertical direction and has a predetermined clearance in a vertical direction in a state of being fastened to the support member by the fastening member. 7. The connector of claim 1, whereinthe steam generator connecting portion further includes: a steam generator sealing member that enables sealing with the outlet of the steam generator. 8. A connector for a steam generator disposed between a steam generator and a flow mixing header and fastening the steam generator to the flow mixing header in a sealing manner, the connector comprising:a base plate mounted on the flow mixing header and having a through hole formed at the center thereof;a steam generator connecting portion protruding along the circumference of the through hole in the base plate and allowing an outlet of the steam generator to be inserted and fastened thereto; andthe steam generator connecting portion has an inner diameter greater than a diameter of the through hole to form a step portion with the through hole, and the outlet of the steam generator is supported by the step portion. 9. The connector of claim 8, whereinthe through hole is formed to be tapered to have a diameter reduced toward an inlet of the flow mixing header from the steam generator connecting portion, and a diameter of the steam generator connecting portion is greater than an inner diameter of the outlet of the steam generator. 10. An integral nuclear reactor comprising:a reactor vessel;an internal structure disposed inside the reactor vessel and dividing the inside of the reactor vessel into an inner space and an outer space to allow a coolant to flow therein;a plurality of steam generators disposed in the outer space of the reactor vessel;a flow mixing header disposed in the outer space of the reactor vessel, allowing the plurality of steam generators to be connected thereto, mixing coolant introduced from the plurality of steam generators, and discharging the mixed coolant; anda connector for a steam generator connecting the steam generator and the flow mixing header and suppressing leakage of a coolant from a connection portion between the steam generator and the flow mixing header, andwherein the connector for the steam generator comprising:a base plate mounted on the flow mixing header and having a through hole formed at the center thereof; anda steam generator connecting portion protruding along the circumference of the through hole in the base plate and allowing an outlet of the steam generator to be inserted and fastened thereto; anda plurality of elastic members provided along the circumference of the steam generator connecting portion on an other surface of the base plate and pressed toward a lower surface of the steam generator so as to be elastically deformed. 11. The integral nuclear reactor of claim 10, wherein the connecting apparatus further comprising:a horizontality adjusting portion provided in plurality along the circumference of the steam generator connecting portion on the other surface of the base plate to support a lower surface of the steam generator, and configured to adjust horizontality of the steam generator. 12. The integral nuclear reactor of claim 11, wherein the horizontality adjusting portion includes:a support protrusion outwardly protruding from the base plate and threaded on an outer circumferential surface thereof; anda height adjusting member fastened to the support protrusion by a screw and moving in a vertical direction by a rotational movement to adjust a height of the steam generator. 13. The integral nuclear reactor of claim 10, wherein the connector, further comprising:an escape preventing portion provided in plurality along the circumference of the steam generator connecting portion on the other surface of the base plate and connected to the lower surface of the steam generator so as not to escape from the steam generator. 14. The integral nuclear reactor of claim 13, wherein the escape preventing portion is connected to the steam generator to have a predetermined clearance. 15. The integral nuclear reactor of claim 14, wherein the escape preventing portion includes: a support member provided as a pair, protruding outwards from the base plate, and allowing a fixing protrusion formed in the steam generator to be inserted therebetween; anda fastening member inserted into the pair of support members and fastening the fixing protrusion to the pair of support members,wherein the fixing protrusion has a slit formed in a vertical direction and has a predetermined clearance in a vertical direction in a state of being fastened to the support member by the fastening member. 16. The integral nuclear reactor of claim 10, wherein the steam generator connecting portion further includes: a steam generator sealing member that enables sealing with the outlet of the steam generator. 17. An integral nuclear reactor comprising:a reactor vessel;an internal structure disposed inside the reactor vessel and dividing the inside of the reactor vessel into an inner space and an outer space to allow a coolant to flow therein;a plurality of steam generators disposed in the outer space of the reactor vessel;a flow mixing header disposed in the outer space of the reactor vessel, allowing the plurality of steam generators to be connected thereto, mixing coolant introduced from the plurality of steam generators, and discharging the mixed coolant; anda connector for a steam generator connecting the steam generator and the flow mixing header and suppressing leakage of a coolant from a connection portion between the steam generator and the flow mixing header, andwherein the connector for the steam generator comprises:a base plate mounted on the flow mixing header and having a through hole formed at the center thereof; anda steam generator connecting portion protruding along the circumference of the through hole in the base plate and allowing an outlet of the steam generator to be inserted and fastened thereto; and wherein the steam generator connecting portion has an inner diameter greater than a diameter of the through hole to form a step portion with the through hole, and the outlet of the steam generator is supported by the step portion. 18. The integral nuclear reactor of claim 17, wherein the through hole is formed to be tapered to have a diameter reduced toward an inlet of the flow mixing header from the steam generator connecting portion, and a diameter of the steam generator connecting portion is greater than an inner diameter of the outlet of the steam generator.
	abstract	An x-ray optical system for producing high intensity x-ray beams. The system includes an optic with a surface formed by revolving a defined contour around a revolving axis that is different than the geometric symmetric axis of the optic. Accordingly, the system may use a source that has a circular emission profile or a large source to provide increased flux to a sample.
	summary
	summary
	abstract	A charged particle beam writing apparatus includes a plurality of tracking calculation units to calculate a deflection amount of the charged particle beam in regard to a movable substrate, a switching unit for each of a plurality of virtual small regions of the substrate, to input an end signal indicating completion of charged particle beam emission to a respective small region, and to switch from output of one of the tracking calculation units to output of another of the tracking calculation units, and a deflector, while a substrate is moving, to deflect the charged particle beam to an n-th small region, based on an output from one of the tracking calculation units before switching and to deflect the charged particle beam to an (n+1)th small region based on an output from another of tracking calculation units after switching the plurality of tracking calculation units.
	abstract	In a method of interactive manipulation of the dose distribution of a radiation treatment plan, after an initial candidate treatment plan has been obtained, a set of clinical goals are transferred into a set of constraints. Each constraint may be expressed in terms of a threshold value for a respective quality index of the dose distribution. The dose distribution can then be modified interactively by modifying the threshold values for the set of constraints. Re-optimization may be performed based on the modified threshold values. A user may assign relative priorities among the set of constraints. When a certain constraint is modified, a re-optimized treatment plan may not violate those constraints that have priorities that are higher than that of the modified constraint, but may violate those constraints that have priorities that are lower than that of the modified constraint.
	claims	1. A method for performing a photo reaction on a target object, comprising:emitting from a light source a continuous output of irradiance onto the target object having a first photo-reactive Zone A, and a second photoreactive Zone B;superimposing over the continuous light output a pulsed light output of irradiance onto the target object, with the pulsed output having a peak irradiance level greater than that of the continuous light output;wherein the target object has a first Zone A that has a photo-reactive character for curing that is different from a Zone B, and the pulsed light output enhances a photo-reaction in Zone A relative to the continuous light output. 2. The method of claim 1 wherein the light source comprises an array of solid state light emitters having a power output/cm2 from about 500 mQ/cm2 to about 4 W/cm2. 3. The method of claim 2 wherein the wavelength of light from the first and second channels is within the range of from about 420 nm to about 150 nm. 4. The method of claim 2 wherein the continuous emission has an irradiance within the range of from about 100 mW/cm2 to about 1.5 W/cm2. 5. The method of claim 4 wherein the pulsed emission has a peak irradiance within a range of from about 1 W/cm2 to about 30 W/cm2 and the ratio for the peak of the pulse to the base of the pulse is at least about 0.02 to about 100. 6. The method of claim 2 wherein the output is provided by groups of emitters comprising a first group of one or more emitters dedicated to the continuous output and a second group of emitters dedicated to providing the pulsed output. 7. The method of claim 2 wherein the output is provided by a group of emitters dedicated to providing both the continuous and pulsed outputs. 8. The method of claim 2 wherein the continuous and pulsed output profiles, respectively, are selected so as to couple into respective absorption bands for predetermined Zone A and Zone B photo-reactants. 9. The method of claim 2 wherein the polymerization reagents in the target object comprise a photo-initiator compound and a monomer or oligomer reactive with the compound. 10. The method of claim 9 wherein the polymerization reagents are selected from the group consisting of reagents for epoxies, acrylates, polyimides and polyamides. 11. The method of claim 9 wherein the method comprises an application fur curing inks; automotive coatings; industrial coatings' cement or concrete coatings; adhesives; tape release polymers in semiconductor processes; paints; or lithography. 12. The method of claim 2 wherein the method is used in an application comprising cleaning or sterilizing the target object. 13. The method of claim 1 wherein the wavelength of light from the first and second channels is within the range of from about 1800 nm to about 150 nm. 14. The method of claim 1 wherein the continuous emission has an irradiance within the range of from about 100 mW/cm2 to about 10,000 mW/cm2. 15. The method of claim 14 wherein the pulsed emission has a peak irradiance within a range of from about 5 W/cm2. 16. The method of claim 1 wherein the duty cycle for the pulsed emission is from about 1% to about 50% at about 0.02 Hz to about 40,000 Hz. 17. The method of claim 1 wherein the duty cycle for the pulsed emission is from about 1% to about −25% at about 0.02 Hz to about 40,000 Hz. 18. The method of claim 1 wherein the duty cycle for the pulsed emission is from about 1% about 10% at about 0.02 Hz to about 40,000 Hz. 19. The method of claim 1 further comprising introducing during light output a laminar flow across a surface area of the material exposed to the light. 20. The method of claim 19 wherein an inerting agent is introduced via the laminar flow. 21. The method of claim 1 wherein the target object comprises CD or DVD. 22. A product comprising a photo-reacted target object, the target object being formed by a method comprising:emitting from a light source a continuous level of irradiance onto the target object having a first photo-reactive Zone A, and a second photoreactive Zone B;superimposing over the continuous light output a pulsed light output having a peak irradiance level greater than that of the irradiance level of the continuous light output;wherein the target object has a first Zone A that has a photo-reactive character for curing that is different from a Zone B, and the pulsed light output enhances a photo-reaction in Zone A relative to the continuous light output.
	summary
	abstract	An improved personal radiation protection system that substantially contours to an operator's body is suspended from a suspension means. The garment is operable to protect the operator from radiation. The suspension means is operable to provide constant force and allows the operator to move freely in the X, Y and Z planes simultaneously, such that the protective garment, face shield, or other attachments integrated into the system are substantially weightless to the operator. The suspension means may be mounted to the ceiling, a vertical wall, the floor, or on a mobile platform. The suspension means may comprise an articulating arm, a balance arm, or a manipulator, and the radiation protection system is suspended generally about its center of gravity.
062360558	summary	BACKGROUND OF THE INVENTION The present invention generally pertains to irradiation systems that utilize a conveyor system for transporting articles through a target region scanned by radiation from a radiation source and is particularly directed to an improvement in positioning the radiation shielding material of the system. A prior art irradiation system that utilizes a conveyor system for transporting articles through a target region is described in U.S. Pat. No. 5,396,074 to Peck et al. In such prior art system, the radiation source and a portion of the conveyor system are disposed in a chamber defined by concrete walls, wherein such concrete walls and additional concrete walls defining an angled passageway into the chamber for the conveyor system shield loading and unloading areas located outside of the chamber from radiation derived from the radiation source. SUMMARY OF THE INVENTION The present invention provides an article irradiation system, comprising a radiation source positioned for scanning a target region with radiation; a conveyor system including a process conveyor positioned for transporting articles in a given direction through the target region; radiation shielding material defining a chamber containing the radiation source, the target region and a portion of the conveyor system; wherein the radiation source is disposed along an approximately horizontal axis inside a loop defined by a portion of the conveyor system and is adapted for scanning the articles being transported through the target region with radiation scanned in a plane transverse to the given direction of transport by the process conveyor; and an intermediate wall of radiation shielding material positioned within the loop and transverse to said approximately horizontal axis. The intermediate wall supports a ceiling of the chamber, inhibits photons emitted from a beam stop disposed in a given wall of the chamber from impinging upon at least one other wall of the chamber and restricts flow throughout the chamber of ozone derived in the target region from the radiation source. Additional features of the present invention are described with reference to the detailed description of the preferred embodiments.
	claims	1. A method comprising attenuating, while hunting for or observing non-human wildlife, one's own emanated electromagnetic field by wearing at least one article of apparel that includes an electromagnetically shielding fabric, which shielding fabric comprises a substantially continuous system of conductive fibers combined with a non-conductive fabric and attenuates the emanated electromagnetic field at frequencies less than about 1 gigahertz, wherein said attenuating of one's own emanated electromagnetic field at frequencies less than about 1 gigahertz decreases the likelihood of detection by a non-human animal. 2. The method of claim 1 wherein the conductive fibers are intermingled with non-conductive fibers that form the non-conducting fabric. 3. The method of claim 1 wherein the conductive fibers are applied to a surface of the non-conducting fabric. 4. The method of claim 1 wherein at least one said article of apparel comprises an article of clothing, footwear, headwear, or eyewear. 5. The method of claim 1 wherein at least one said article of apparel includes a visual camouflage pattern on at least a portion of its outer surface. 6. The method of claim 1 wherein at least one said article of apparel includes an odor absorber, suppressant, attenuator, or blocker. 7. The method of claim 1 wherein the shielding fabric includes between about 2% and about 35% by weight of the conductive fibers. 8. The method of claim 1 wherein the conductive fibers comprises stainless steel fibers. 9. The method of claim 1 wherein the conductive fibers comprise copper, silver, conductive ceramic, conductive polymer, or conductive nanotubes. 10. A method comprising attenuating, while a user is hunting for or observing non-human wildlife, the user's emanated electromagnetic field by:providing to the user at least one article of apparel that includes an electromagnetically shielding fabric, which shielding fabric comprises a substantially continuous system of conductive fibers combined with a non-conductive fabric and attenuates the emanated electromagnetic field at frequencies less than about 1 gigahertz; andinstructing the user to wear, while hunting for or observing the non-human wildlife, at least one said article of apparel,wherein said attenuating of the user's emanated electromagnetic field at frequencies less than about 1 gigahertz decreases the likelihood of detection of the user by a non-human animal. 11. The method of claim 10 wherein the conductive fibers are intermingled with non-conductive fibers that form the non-conducting fabric. 12. The method of claim 10 wherein the conductive fibers are applied to a surface of the non-conducting fabric. 13. The method of claim 10 wherein at least one said article of apparel comprises an article of clothing, footwear, headwear, or eyewear. 14. The method of claim 10 further comprising constructing at least one said article of apparel prior to providing it to the user. 15. The method of claim 10 wherein at least one said article of apparel includes a visual camouflage pattern on at least a portion of its outer surface. 16. The method of claim 10 wherein at least one said article of apparel includes an odor absorber, suppressant, attenuator, or blocker. 17. The method of claim 10 wherein the shielding fabric includes between about 2% and about 35% by weight of the conductive fibers. 18. The method of claim 10 wherein the conductive fibers comprises stainless steel fibers. 19. The method of claim 10 wherein the conductive fibers comprise copper, silver, conductive ceramic, conductive polymer, or conductive nanotubes. 20. The method of claim 1 wherein said attenuating of one's own emanated electromagnetic field at frequencies less than about 1 megahertz decreases the likelihood of detection by the non-human animal. 21. The method of claim 1 wherein said attenuating of one's own emanated electromagnetic field at frequencies less than about 1 kilohertz decreases the likelihood of detection by the non-human animal. 22. The method of claim 10 wherein said attenuating of the user's emanated electromagnetic field at frequencies less than about 1 megahertz decreases the likelihood of detection of the user by the non-human animal. 23. The method of claim 10 wherein said attenuating of the user's emanated electromagnetic field at frequencies less than about 1 kilohertz decreases the likelihood of detection of the user by the non-human animal.
	summary
	description	Referring now to the drawings in detail and in particular to FIG. 1 there is generally illustrated a boiling water reactor 10 of a commercial nuclear power plant for generating electricity. The nuclear reactor 10 is generally characterized by a reactor pressure vessel 12 and by a RRS 14. The RRS has two substantially identical recirculation loops A and B hydraulically connected in parallel with each other and in series with the reactor pressure vessel 12. During on-line electrical power generation, coolant water (high purity water containing parts per million or lower levels of various ions and, in some cases, dissolved hydrogen gas) is pumped by feedwater pumps (not shown) from a turbine-generator unit (not shown) in a closed loop into the reactor pressure vessel 12 through an inlet nozzle 16 and steam is generated within the reactor pressure vessel 12. The steam flows out of the pressure vessel 12 through an outlet nozzle 18 back to the turbine-generator unit. The reactor recirculation loops A and B provide high velocity coolant water to a plurality of jet pumps 20 located within the reactor pressure vessel 12 for facilitating the flow of coolant water within the pressure vessel 12. As FIG. 1 shows, the reactor pressure vessel 12 includes a bottom head 22 with a sidewall 24 extending vertically to a flange 26. A removable head 28 has a flange 30 that may be bolted to the reactor pressure vessel flange 26. The reactor pressure vessel 12 has a core shroud 32 and a core plate 34, which define a central core region for containing removable fuel assemblies 38. The core shroud 32 has a removable upper end 40 that may be removed in order to remove the fuel assemblies 38. The pressure vessel wall 24, core shroud 32 and ring member 42 define an annulus region surrounding the central core region. The reactor pressure vessel bottom head 22 and the core plate 34 define a lower internals region which is in fluid flow communication with the central core region via flow holes 44 in the core plate 34. As FIG. 1 also shows, each reactor circulation loop A,B of the reactor circulation system 14 generally includes a centrifugal pump 50 having a pump suction nozzle and a pump discharge nozzle. The pumps 50 of commercial boiling water reactors may have nominal capacities of up to about 100,000 gallons per minute or more and the pump nozzles may have diameters of up to about 28 inches or more. Each pump suction nozzle is connected by suction piping 54 with a nozzle 52 in the pressure vessel wall 24 for fluid flow connection with the annulus region of the pressure vessel 12. In another commercial boiling water reactor design (not shown), the vessel nozzle 52 may connect the suction piping 54 with the lower internals region of a reactor pressure vessel. The discharge nozzle of each centrifugal pump 50 is connected by discharge piping 56 to a plurality of reactor pressure vessel nozzles, illustrated in FIG. 1 by nozzle 58. As is illustrated in FIG. 3, the discharge piping 56 comprises at least one manifold 64 with pipe reducers 66 which divide each loop A,B into parallel branches 68. In addition, the branches 68 may further subdivide each loop A,B into parallel pipes 70. In any event, each pipe 70 extends to one of the vessel nozzles 58. As is best shown in FIG. 2, each reactor pressure vessel nozzle 58 extends to a riser pipe, illustrated by riser pipe 72, in the annulus region of the reactor pressure vessel 12. Each riser pipe 72 extends upwardly to a ram""s head manifold 74 having two 180xc2x0 piping bends 76 disposed in parallel adjacent a pair of downstream jet pumps 20. It should be noted that FIG. 2 shows a jet pump assembly that has been modified to facilitate a full loop decontamination during a scheduled outage in accordance with the practice of the present invention. As shown, the piping bends 76 have been sealed with plugs 86, the jet pump inlet sections 78 (shown in FIG. 1) of the jet pumps 20 have been removed and the remaining portions 80 of the jet pumps 20 have been capped. After generating electric power during on-line operations for a year or more, commercial nuclear reactors exemplified by nuclear reactor 10 are taken off-line for refueling and/or performing scheduled maintenance or repairs. It is often desirable to first decontaminate the recirculation loops A and B but not the internal regions of the reactor pressure vessels 12 in order to reduce the radiation levels in the containment building before performing the scheduled maintenance or repairs. In a preferred practice in accordance with the present invention, a boiling water reactor 10 having a reactor pressure vessel 12 with two recirculation loops A,B piped thereto is decontaminated by: installing plugs 86 on both outlets of the jet pump ram""s head manifolds 74 in both of the recirculation loops A,B; pumping a decontamination solution in the loops A,B to wash the irradiated oxide layers on the piping surfaces; introducing compressed air, nitrogen or other monitoring gas into selected ram""s head manifolds 74 in the loops A,B through one of the plugs 86 for monitoring the pressure of the process in the manifolds; and determining the level of the decontamination solution in the recirculation loops A,B from the pressure of the monitoring gas in the manifolds 74. The preferred practice of the present invention may also include the additional steps of installing additional plugs 84 in the reactor vessel nozzles 52 connected with the suction pipes 54 extending to the suction connections of the recirculation pumps 50 in the recirculation loops A,B before pumping the decontamination solution into the loops A,B; introducing a monitoring gas through the plugs 84 into the suction pipes 54 for monitoring the monitoring gas pressure in the suction pipes 54; and determining the level of the decontamination solution in the suction pipes 54 from the pressure the monitoring gas introduced through the plugs 84. FIGS. 2 and 3 generally shows the recirculation loops A,B isolated from the reactor pressure vessel 12 by plugs 84 installed in the pressure vessel nozzles 52 connected with the suction piping 54 and by other plugs 86 installed in the outlets of the 180xc2x0 bends 76 of the ram""s head manifolds 74. As FIG. 3 illustrates, the pressure vessel nozzles 52 and their installed plugs 86 are located below the ram""s head manifolds 74 and their installed plugs 86. In commercial boiling water nuclear reactors, the pressure vessel nozzles 52 may be up to about ten feet or more below the ram""s head manifolds 74. The plugs 84 and 86 may be round aluminum or stainless steel plugs with inflatable rubber bladders. The plugs 84 and 86 may be installed by tooling operated by robots (not shown) working in the annulus region of the pressure vessel 12 after the reactor pressure vessel head 28 has been removed. It should be noted that the plugs 84,86 are installed while the reactor pressure vessel 12 is submerged in water. Accordingly, the loops A,B are solid with water after they have been isolated. Also, the loops A,B may be isolated from the nuclear reactor""s residual heat removal system by closing the valves in piping 92. In the preferred decontamination practice, a pumping unit such as a skid mounted decontamination unit 100 having a pump 102 as shown in FIG. 3 may be connected with the loops A,B at decontamination flanges 94 located near the suction connections of the recirculation pumps 50 for pumping the decontamination solution in the loops A,B. It is to be noted at this point that the decontamination solution is usually pumped through the recirculation pumps 50, which are not used because they are too big to be useful during a decontamination process. Flexible piping sections 104 may be used to facilitate the temporary connection of piping extending from the decontamination unit 100 with the decontamination flanges 94 of the loops A,B. In addition, the decontamination unit 100 may have valves 106 and associated piping that will permit the skid mounted pump 102 to pump the decontamination solution from one loop A,B to the other loop A,B. The arrangement of valves 106 shown in FIG. 3 advantageously permits the flow of decontamination solution to be reversed as often as is desired. In addition, in other practices of the present invention, jumper connections (not shown) may be employed to connect the manifolds 86 of one loop A,B with the manifolds 86 of the other loop A,B so that the skid mounted pump 102 may be employed to continuously recirculate the decontamination solution through the loops A,B. In the preferred practice of the present invention, both loops A,B may be isolated by plugging as described above and then drained down to an intermediate level. For example, the loops A,B may be drained down to a level where the RRS 14 is about half full. The volume in the loops A,B between the plugs 84, 86 and the fluctuating liquid levels of the decontamination solution in the loops A,B may be back filled by air, nitrogen or other suitable gas and vented through connections in the plugs 84,86 described below. Preferably, the pressure above the liquid levels of the decontamination solution throughout the RRS 14 are substantially equalized and nominally about atmospheric pressure throughout the decontamination process in this practice. As shown in FIG. 3, the decontamination unit 100 may also have a feed tank 110 and a feed pump 112 for feeding a suitable decontamination solvent or mixture of solvents (preferably diluted as an aqueous solution) to the suction side of the skid mounted pump 102. The skid mounted pump 102 may then mix the solvent with the coolant water within the pump body and pump the diluted solvent and at least some of the coolant water in one of the two loops A,B into the other of the two loops A,B, including the riser pipes 72. Thus, for example, the pump 102 first may pump some of the coolant water from loop A and the diluted solvent from feed tank 110 into loop B. Later, the valves 106 may be reversed and the skid mounted pump 102 may pump the decontamination solution from the loop B back to loop A. Advantageously, the energy input from the skid mounted pump 102 may also cause the decontamination solution to slosh and splash against the oxide layers on the surfaces of the piping of the recirculation loops A,B and the riser pipes 72. As is also shown in FIG. 3, the skid mounted unit 100 may have a heater 114 for heating the decontamination solution up to a desired operating temperature for effectively dissolving the oxide layers in reasonable time periods. For example, known dilute chemical decontamination processes may be performed at temperatures up to about 150xc2x0 F. or 180xc2x0 F. The skid mounted unit 100 may also have ion exchangers represented by ion exchanger 116 for cleaning up the decontamination solution at the end of the decontamination process to remove the decontamination solvents and dissolved ions from the coolant water. The levels of the decontamination solutions in the loops A,B must be known at all times during the course of decontamination processes to effectively decontaminate the loops A,B without running the risk of pumping the decontamination solvents into the reactor pressure vessels 12. In addition, it is often desirable that the solvents in the decontamination solutions not contact plugs 84 (and particularly aluminum plugs) in the pressure vessel nozzles 52. In the practice of the present invention, commercial gas bubbling level/pressure monitoring systems (not shown) are used for monitoring the levels of the decontamination solutions in the loops A,B. These monitoring systems actually determine the level of a process liquid at a particular location by sensing a differential pressure due to the vertical height (or static head) of the liquid above a reference level and then calculating the level based upon the differential pressure and the specific gravity of the liquid. Commercially available monitoring devices sense the pressures at various locations in a process by introducing (or, as it is sometimes known in the field, xe2x80x9cbubblingxe2x80x9d) compressed air or nitrogen at known, low flow rates into the process and sensing the pressure of the monitoring gas. Each sensed pressure is converted to an appropriate signal by a transducer and the signal is sent to a programmable logic controller or other suitable calculating device by a pressure transmitter. In the preferred practice of the present invention, the monitoring systems are employed to sense the pressure within the loops A,B at the plugs 84, 86 and at the decontamination flanges 88 on the suction sides of the recirculation pumps 50, which flanges 88 are physically located low in the drywells of nuclear reactors. Additional monitoring systems may be optionally employed to sense the pressures at other locations. Advantageously, the monitoring systems may be located remotely from the pressure vessel 12 in relatively low radiation level areas, e.g., on the refuel floor, and connected with one or both recirculation loops A,B via small bore monitoring gas sensing lines 120. The monitoring gas sensing lines 120 may extend through the one of the plugs 86 in the bends 76 of the manifolds 74 for introducing the monitoring gas into the manifolds 74 and associated riser pipes 72. Although it may be sufficient to introduce the monitoring gas into only one of the manifolds 74 of one of the loops A,B to determine the level of the decontamination solution, it is preferred to introduce the monitoring gas into at least two of the manifolds 74 in each loop A, B in order to assure the accuracy of the measurement. Similarly, monitoring gas sensing lines 122 may extend through the plug 84 in the reactor pressure vessel nozzle 52 connected with the suction piping 54 for sensing the pressure of the decontamination solution in the suction piping 54 extending to the suction connection of the recirculation pumps 50. As is shown by FIG. 3, the ends of the monitoring gas sensing lines 120, 122 preferably have flexible lengths 104 near the connections with the plugs 86 and 84, respectively. In the preferred practice of the present invention, the changing gas volumes above the surface levels of the decontamination solution in the loops A,B adjacent the plugs 84,86 and suction pipes 54 are vented by vent lines 130 to reduce the effects of process changes that might substantially affect the back pressure on the monitoring gas sensing lines 120,122 and thereby result in inaccurate sensed pressures. Preferably, for the manifolds 86 that are connected with the monitoring gas sensing lines 120, the monitoring gas is introduced through one plugged outlet of each such manifold 74 and the monitoring gas is vented through the other plugged outlet of the manifold. The vent lines 130 may be small bore hoses, piping or tubing of about xe2x85x9c inch diameter or larger. As is shown in FIG. 3, the vent lines 130 from the recirculation loops A and B are preferably directed to one of two vent tanks 132 and the vent tanks 132 are vented through vent lines 134 to a vacuum tank 136. Advantageously, this arrangement permits the overpressure on the decontamination solution in the several portions of the loops A,B to be substantially equalized as the decontamination solution is pumped from loop A,B to loop A,B in the course of a decontamination process. The vacuum tank 136 may be connected with a vacuum pump 138 for pulling a vacuum on the vacuum tank 136 and on the vent lines 130 and 134. Advantageously, the vacuum may be employed to clear any trapped water in the 180xc2x0 bends 76 of the ram""s head manifolds 74. It should be noted that the structure of the 180xc2x0 bends 76 will inherently trap coolant water on the plugged outlet side at the time the plugs 84,86 are installed. This water will remain trapped even though the level of the coolant water throughout the loops A,B is lowered. Later, during the decontamination process, the decontamination solution may flow over the central portion of the bends 76 and become trapped in the plugged outlets. The accumulation of water in the plugged outlets may be indicated by excessively high indicated pressure levels. In the preferred practice, the vacuum may be periodically employed when the indicated pressure indicates that the vent paths have become blocked. Also, the vent tanks 132 preferably have drain pipes 140 extending to the vacuum tank 136 for draining coolant water and decontamination solution which may have vented into the vent tanks 132. An eductor 142 or other pumping means may be connected by pipe 144 to the vacuum tank 136 for pumping decontamination solution out of the vacuum tank 136. The vent tanks 132, vacuum tank 136, vacuum pump 138 and eductor 142 may be located on the skid mounted unit 100 together with the monitoring system on the refuel floor or at another suitable location remote from the reactor pressure vessel 12. The ram""s head manifolds 74 and the vent tanks 110 may be drained on an xe2x80x9cas neededxe2x80x9d basis whenever the solution begins to block the vent path. While a present preferred embodiment of the present invention has been shown and described, it is to be understood that the invention may be otherwise variously embodied within the scope of the following claims of invention.
	claims	1. A system for irradiating articles, including, at least first, second and third process conveyor segments disposed in spaced relationship to one another for moving the articles, there being a gap between the first and second process segments and a gap between the second and third process segments, each of the process conveyor segments being movable at a speed consistent with the speeds of the other process conveyor segments, a first radiation source disposed relative to the articles on the process conveyor segments, in the gap between the first and second process conveyor segments, for directing radiation in a first direction to the articles, a second radiation source disposed relative to the articles on the process conveyor segments, in the gap between the second and third process conveyor segments, for directing radiation to the articles in a second direction opposite to the first direction, the second radiation source being displaced from the first radiation source in the direction of the movement of the articles on the process conveyor segments, a loading area for the articles, the loading area being disposed before the process conveyor segments in the direction of movement of the articles on the process conveyor segments, and an unloading area for the articles, the unloading area being disposed after the process conveyor segments in the direction of movement of the articles on the process conveyor segments. 2. An irradiation system as set forth in claim 1 , including, claim 1 a divider in the process conveyor segments for producing two spaced and parallel tracks on the process conveyor segments, each of the tracks being constructed to transport the articles on the track separate from the articles on the other track but simultaneously with the articles on the other track. 3. A radiation source as set forth in claim 2 , including, claim 2 each of the tracks in each of the process conveyor segments including rollers disposed to move the articles on the track toward the articles on the other track, during the movement of the articles toward the first and second radiation sources, to obtain a disposition of the articles on the first and second tracks in substantially contiguous relationship to each other at the positions of the first and second radiation sources, each of the first and second radiation sources being provided with a width corresponding substantially to the combined widths of the articles on the two (2) spaced tracks. 4. A system as set forth in claim 3 wherein claim 3 the process conveyor segments are disposed in a horizontal plane and wherein one of the radiation sources is disposed above the articles on the process conveyor segments and points downwardly toward the articles on the process conveyor segments and wherein the other of the radiation sources is disposed below the articles on the process conveyor segments and points upwardly toward the articles on the process conveyor segments. 5. A system for irradiating articles disposed in article carriers, including, a process conveyor having first, second and third segments disposed in a series relationship and separated from one another to define first and second gaps each having a length less than the width of the articles, a first load conveyor for transporting the articles and for transferring the articles to the first segment in the process conveyor, a second load conveyor for receiving the articles from the third segment in the process conveyor and for transporting the articles, a first radiation source disposed relative to the first gap for directing radiation toward the first gap in a first direction toward the articles on the process conveyor, and a second radiation source disposed relative to the second gap for directing radiation toward the second gap toward the articles on the process conveyor in a second direction opposite to the first direction. 6. A system as set forth in claim 5 wherein claim 5 the cumulative radiation from the first and second sources at each position in the articles on the process conveyor is between first and second limits. 7. A system as set forth in claim 5 wherein claim 5 the load conveyors and the process conveyor have dividers for producing a pair of parallel tracks each for transporting articles between the loading area and the unloading area at the same time that the other track transports articles between the loading area and the unloading area. 8. A system as set forth in claim 7 , including, claim 7 structure for diverging the articles on each of the tracks in the first load conveyor from the divider during the movement of the articles on the first load conveyor toward the process conveyor. 9. A system as set forth in claim 8 , including, claim 8 structure for converging the articles on each of the tracks toward the divider during the movement of the articles on the process conveyor toward the first radiation source. 10. A system as set forth in claim 9 , including, claim 9 structure for maintaining the convergence of the articles on each of the tracks on the process conveyor during the movement of the articles on the process conveyor from the first radiation source toward the second radiation source. 11. A system as set forth in claim 6 , including, claim 6 the load conveyors and the process conveyor having dividers for producing a pair of parallel tracks each for transporting articles between the loading area and the unloading area at the same time that the other track transports articles between the loading area and the unloading area, structure for diverging the articles on each of the tracks on the first load conveyor from the divider during the movement of the articles on the first load conveyor toward the process conveyor, structure for converging the articles on the process conveyor toward the divider during the movement of the articles on the process conveyor toward the first radiation source, and structure for maintaining the convergence of the article carriers on each of the tracks of the process conveyor during the movement of the articles on the process conveyor from the first radiation source toward the second radiation source. 12. A system for irradiating articles, including, a radiation source, a process conveyor for moving the articles past the radiation source for an irradiation of the articles by the radiation source, a load conveyor disposed relative to the process conveyor for transferring the articles to the process conveyor at a speed for movement of the articles on the process conveyor, a divider on the load conveyor for dividing the load conveyor into a pair of parallel tracks each constructed to transport individual ones of the articles at the same time as the transport of other ones of the articles on the other track, and members disposed on the tracks for converging the articles on the tracks toward one another for movement of the articles in the converged relationship on the process conveyor past the radiation source. 13. A system as set forth in claim 12 wherein claim 12 the members constitute rollers rotatable in a direction to advance the articles toward the radiation source and to converge the articles during the advance of the articles toward the radiation source. 14. A system as set forth in claim 12 wherein claim 12 the radiation source has a width to irradiate the articles on the tracks in the converged relationship of the articles on the tracks on the process carrier. 15. A system as set forth in claim 12 wherein claim 12 the members constitute first members and wherein the first members are disposed on the process conveyor before the radiation source in the direction of movement of the articles on the tracks to converge the articles and wherein second members are disposed on the load conveyor on the tracks at a position after the movement of the articles past the radiation source for diverging the articles on each of the tracks away from the articles on the other track. 16. A system as set forth in claim 14 wherein claim 14 the divider is disposed to define the tracks at the position of the diverging relationship of the articles on the tracks on the load conveyors and at the position of the converging relationship of the articles on the tracks on the process conveyor. 17. A system for irradiating articles, including, a radiation source, a process conveyor for moving the articles past the radiation source for an irradiation of the articles by the radiation source, a load conveyor disposed relative to the process conveyor for transferring the articles to the process conveyor at a speed for movement of the articles on the process conveyor, a divider on the load conveyor for dividing the load conveyor into a pair of parallel tracks each constructed to transport articles at the same time as the transport of articles on the other track, and members disposed on the tracks on the load conveyor for diverging the articles on each of the tracks on the load conveyor away from the articles on the other track on the load conveyor for movement of the articles in the diverged relationship on the load conveyor. 18. A system as set forth in claim 17 wherein claim 17 the members constitute rollers rotatable in a direction to advance the articles toward the radiation source and away from the divider. 19. A system as set forth in claim 17 wherein claim 17 the divider is disposed on the load conveyor to define the tracks at the position of the diverging relationship of the articles on the tracks. 20. A system for irradiating first and second opposite sides of articles, including, a radiation source, a first load conveyor divided into first and second tracks and disposed before the radiation source in the direction of movement of the articles and constructed to move the articles on the tracks, a process conveyor responsive to the movement of the articles on the first load conveyor for receiving the articles from the first load conveyor, the process conveyor being operative to move the articles past the radiation source for an irradiation of the first sides of the on the process conveyor by the radiation source, a second load conveyor divided into first and second tracks and disposed after the radiation source in the direction of the movement of the articles and constructed to receive on its first and second tracks the articles respectively disposed on the first and second tracks of the process conveyor after the irradiation of the articles on the process conveyor by the radiation source, the radiation source being constructed to initially irradiate the articles on the first track of the process conveyor and to subsequently irradiate the articles on the second track of the process conveyor, a third process conveyor coupling the first track on the second load conveyor and the second track on the first load conveyor for transferring the articles from the first track on the second load conveyor to the second track on the first load conveyor, after the irradiation by the source of the articles on the first track of the process conveyor, for movement of the articles past the radiation source a second time for irradiation of the second sides of the articles on the process conveyor by the radiation source, and a device coupled to the third load conveyor for inverting the articles transferred to the third load conveyor from the first track in the second load conveyor to obtain the irradiation of the second sides of the articles by the radiation source. 21. A system as set forth in claim 20 , including, claim 20 each of the first, second and third load conveyors being formed from rollers at progressive positions along the load conveyors and the first and second tracks on each of the first and second load conveyors being defined by dividers extending along the load conveyors in the direction of movement of the articles on the load conveyors. 22. A system as set forth in claim 20 , including, claim 20 a loading area disposed relative to the first load conveyor for providing for a transfer of the articles from the loading area to the first track on the first load conveyor, and an unloading area disposed relative to the second track on the second load conveyor for receiving the articles after the irradiation of the first and second sides of the articles by the radiation source. 23. A system as set forth in claim 19 wherein claim 19 the cumulative radiation from the first and second radiation sources at each position in the articles is between first and second limits. 24. A system as set forth in claim 20 , including, claim 20 a loading area disposed relative to the first load conveyor for providing for a transfer of the articles from the loading area to the first track on the first load conveyor, and an unloading area disposed relative to the second track on the second load conveyor for receiving the articles after the radiation of the first and second sides of the articles by the radiation source, the cumulative radiation from the first and second radiation sources at each position in the articles being between first and second limits, and the process conveyor being constructed to convey the articles on the process conveyor at a substantially constant speed past the radiation source. 25. A system for irradiating articles, including, a load conveyor divided into first and second tracks and disposed to receive individual ones of the articles on the first track and simultaneously to receive others of the articles on the second track, a process conveyor disposed relative to the first load conveyor for receiving the articles from the first and second tracks on the first load conveyor and having first and second tracks and constructed to receive on the first and second tracks the articles respectively from the first and second tracks on the first load conveyor, the process conveyor being constructed to move the articles on the process conveyor at a particular speed and being provided with first and second gaps spaced from each other in the direction of movement of the articles on the process conveyor, a first radiation source disposed at the first gap for directing radiation toward the first gap and through the articles on the process conveyor in a first direction transverse to the direction of movement of the articles on the process conveyor and through a width encompassing the articles on the first and second tracks of the process conveyor, and a second radiation source disposed at the second gap for directing radiation toward the second gap and through the articles in a second direction opposite to the first direction and through a width encompassing the articles on the first and second tracks of the process conveyor. 26. A system as set forth in claim 25 , including, claim 25 a loading area disposed relative to the load conveyor for providing for a transfer of the articles from the loading area to the first and second tracks of the load conveyor, the load conveyor constituting a first load conveyor, a second load conveyor having first and second tracks for respectively receiving the articles from the first and second tracks of the process conveyor after the irradiation of the articles on the first and second tracks of the process carrier by the first and second radiation sources, and unloading area for receiving the article carriers on the first and second tracks of the second load conveyor. 27. A system as set forth in claim 25 , including, claim 25 the articles having first and second opposite sides, a third load conveyor having a track extending from the first track of the second load conveyor to the second track of the first load conveyor, an inverter on the third load conveyor for inverting the article carriers on the third load conveyor, and a controller responsive to a failure of one of the first and second radiation sources for activating the third load conveyor to receive the article carriers on the first track of the second load conveyor and for activating the inverter to provide for an inversion of the articles on the third load conveyor and for transferring the inverted articles to the second track on the first load conveyor for an irradiation of the second side of the articles by the one of the first and second radiation sources. 28. A system as set forth in claim 25 , including, claim 25 the articles having first and second opposite sides, and the controller being associated with the first and second load conveyors for providing for an irradiation of the second sides of the articles by one of the first and second radiation sources, after the irradiation of the first sides of the articles by the one of the radiation sources, when the other one of the radiation sources is unable to provide such irradiation of the second sides of the articles. 29. A system as set forth in claim 27 wherein claim 27 the controller provides for an inversion of the articles after the irradiation of the first sides of the articles by the one of the radiation sources and provides for another movement of the articles past the one of the radiation sources after the inversion of the articles to obtain an irradiation of the second sides of the articles by the one of the radiation sources. 30. A system for irradiating articles disposed on article carriers, including, a first load conveyor, a transport mechanism for advancing the article carriers, a first robotic device for removing the articles individually from the article carriers on the transport mechanism and for transferring the articles to the first load conveyor, a source of radiation, a process conveyor for receiving the articles from the first load conveyor and for moving the articles past the radiation source at a substantially constant speed for an irradiation of the articles by the radiation source, a second load conveyor for receiving the irradiated articles from the process conveyor and for moving the irradiated articles toward the article carriers on the transport mechanism, and a second robotic device for removing the irradiated articles individually from the second load conveyor and for transferring the irradiated articles to the article carriers on the transport mechanism, the articles having first and second opposite sides, the source of radiation constituting a first source of radiation, a second source of radiation, there being first and second gaps in the process conveyor, the first source of radiation being disposed at the first gap in the process conveyor to direct the radiation in a first direction toward the first side of the articles, the second source of radiation being disposed at the second gap in the process conveyor to direct the radiation toward the second side of the articles in a second direction opposite to the first direction. 31. A system as set forth in claim 31 wherein claim 31 apparatus is provided for obtaining a radiation of the first and second opposite sides of the articles on the process conveyor by one of the first and second sources of radiation when the other one of the sources of radiation is unable to irradiate the articles on the process conveyor. 32. A system as set forth in claim 30 , including, claim 30 each of the load conveyors and the process conveyor including first and second tracks each constructed to transport articles at the same time as the transport of articles by the other track, a third load conveyor coupling the first track on the second load conveyor and the second track on the first load conveyor to obtain a transfer of articles from the first track on the second load conveyor to the second track on the first load conveyor after the irradiation of the first side of the articles on the first track of the first load conveyor by one of the sources of radiation when the other one of the sources of radiation fails to irradiate the second side of the articles, and an inverter for inverting the articles during the transfer of the articles from the first track of the second load converter to the second track of the first load converter, the articles being irradiated on the second side of the articles upon the movement of the articles on the second track of the first load conveyor and the transfer of the articles to the process conveyor for movement past the radiation source. 33. A system as set forth in claim 30 wherein claim 30 the source of radiation irradiates the articles on the process conveyor at every position in the articles with a strength between first and second limits. 34. A method of irradiating articles disposed on article carriers, including the steps of: moving the article carriers on a transport mechanism, transferring articles in sequence from the article carriers to a first load conveyor during the movement of the article carriers on the transport mechanism, moving the articles on the first load conveyor to a process conveyor, moving the articles at a substantially constant speed on the process conveyor past sources of radiation to irradiate opposite sides of the articles wherein there are first and second sources of radiation and wherein the first source of radiation irradiates the first opposite side of the articles on the process conveyor and the second source of radiation irradiates the second opposite side of the articles on the process conveyor, transferring the irradiated articles from the process conveyor to a second load conveyor for movement of the irradiated articles to the transport mechanism, and transferring the irradiated articles on the second load conveyor to article carriers on the transport mechanism, and wherein the articles are inverted when one of the sources of radiation fails to irradiate the articles on the process conveyor and wherein xe2x80x83the other one of the sources of radiation irradiates the inverted articles on the process conveyor to obtain an irradiation of the second opposite side of the articles. 35. A method of irradiating articles disposed on article carriers, including the steps of: moving the article carriers on a transport mechanism, transferring articles in sequence from the article carriers to a first load conveyor during the movement of the article carriers on the transport mechanism, moving the articles on the first load conveyor to a process conveyor, moving the articles at a substantially constant speed on the process conveyor past sources of radiation to irradiate opposite sides of the articles, transferring the irradiated articles from the process conveyor to a second load conveyor for movement of the irradiated articles to the transport mechanism, and transferring the irradiated articles to article carriers on the transport mechanism, the first and second load conveyors and the process conveyor are divided to form two parallel tracks and wherein the articles are simultaneously disposed on the first and second tracks of each of the first and second load conveyors and wherein the articles on the two tracks of the first load conveyor are diverged before the transfer of the articles to the process conveyor and wherein the articles on the process conveyor are converged before the irradiation of the articles by the radiation source. 36. A method of irradiating articles disposed on article carriers, including the steps of: moving the article carriers on a transport mechanism, transferring articles in sequence from the article carriers to a first load conveyor during the movement of the article carriers on the transport mechanism, moving the articles on the first load conveyor to a process conveyor, moving the articles at a substantially constant speed on the process conveyor past sources of radiation to irradiate opposite sides of the articles, transferring the irradiated articles from the process conveyor to a second load conveyor for movement of the irradiated articles to the transport mechanism, and transferring the irradiated articles to article carriers on the transport mechanism, the articles are inverted when the one of the sources of radiation fails to irradiate the articles and wherein the other one of the sources of radiation irradiates the inverted articles to obtain an irradiation of the second side of the articles and wherein the first and second load conveyors and the process conveyor are divided to form two parallel tracks and wherein the articles are simultaneously disposed on the first and second tracks of the first and second load conveyors and wherein the articles on the two tracks of the first load conveyor are diverged and wherein xe2x80x83the articles on the process conveyor are converged before the irradiation of the articles by the radiation sources or by the other one of the radiation sources when the one of the radiation sources fails to irradiate the articles. 37. A method of irradiating articles disposed on article carriers, including the steps of: providing a transport mechanism for the article carriers, transferring articles in sequence to a first load conveyor from each of the successive article carriers on the transport mechanism, transporting the articles in sequence at a first speed on the first load conveyor to a process conveyor, moving the articles in sequence on the process conveyor at a substantially constant speed less than the first speed, irradiating first and second opposite sides of the articles on the process conveyor, transferring the irradiated articles to a second load conveyor, providing for the movement of the articles on the second load conveyor at a speed greater than the speed of the articles on the process conveyor, and transferring the articles on the second load conveyor to article carriers on the transport mechanism, and wherein first and second sources of radiation are respectively disposed on first and second opposite sides of the articles disposed on the process conveyor to irradiate opposite sides of the articles and wherein the first side of the articles is irradiated in a first pass of the articles past one of the radiation sources and wherein the second side of the articles is irradiated in a second pass of the articles past the one of the radiation sources when the other one of the radiation sources is not operative to irradiate the articles. 38. A method as set forth in claim 37 wherein claim 37 the articles are inverted in the time between the first and second passes of the articles past the one of the radiation sources. 39. A method of irradiating articles, including the steps of: providing at least one source of radiation, providing a loading area, displaced from the source of radiation, for holding the articles, providing an unloading area displaced from the at least one source of radiation and the loading area, providing for a transfer of the articles from the loading area to a first load conveyor, providing for a transfer of the articles from the first load conveyor to a process conveyor to obtain a movement of the articles past the at least one source of radiation for an irradiation of the articles on the process conveyor by the at least one radiation source, providing for a transfer of the articles from the process conveyor to a second load conveyor for a transport of the articles to the unloading area, providing for a division of the first and second load conveyors and the process conveyor into a pair of tracks each constructed to transport articles at the same time as the transport of articles on the other track, providing for a divergence of the articles on the first and second tracks of the first load conveyor after the transfer of the articles to the first load conveyor from the loading area, and providing for a convergence of the articles on the process conveyor before the movement of the articles on the process conveyor past the at least one radiation source. 40. A method as set forth in claim 39 wherein claim 39 the process conveyor provides for a movement of the articles past the at least one source of radiation at a particular speed and wherein the process conveyor is provided in segments with a gap between each pair of successive segments and wherein the at least one source of radiation constitutes first and second sources of radiation each disposed at an individual one of the gaps in the process conveyor and wherein the first source of radiation is disposed relative to the articles on the process conveyor to irradiate the first side of the articles in a first direction and wherein the second source of radiation is disposed relative to the articles on the process chamber to irradiate the second side of the articles in a second direction opposite to the first direction. 41. A method as set forth in claim 39 wherein claim 39 the at least one source of radiation constitutes first and second sources of radiation and wherein the first source of radiation irradiates the articles in a first direction through the articles from the first side of the articles and wherein the second source of radiation irradiates the articles in a second direction through the articles from the second side of the articles where the second direction is opposite to the first direction and wherein the cumulative amount of irradiation of the articles at every position in the articles by the first and second sources of radiation is between first and second particular limits. 42. A method as set forth in claim 39 wherein claim 39 the first and second opposite sides of the articles constitute the tops and bottoms of the articles and wherein the first and second sources of radiation are respectively disposed above and below the articles on the process conveyor and wherein the cumulative amount of irradiation of the articles on the process conveyor at every position in the articles by the first and second sources of radiation is between first and second particular limits. 43. A method of irradiating articles disposed in a non-uniform relationship on article carriers, including the steps of: providing a loading area, providing an unloading area displaced from the unloading area, providing a transport mechanism movable past the loading and unloading areas with the article carriers disposed on the transport mechanism, providing a process conveyor, providing a controlled transfer of each of the articles in sequence from each of the successive article carriers on the transport mechanism in a substantially uniform relationship relative to the process conveyor for each of the articles, providing a movement of the articles on the process conveyor past at least one source of radiation to obtain an irradiation of the articles on the process conveyor, and providing a transfer of the irradiated articles to the article carriers on the transport mechanism as the article carriers move on the transport mechanism past the unloading area wherein the process conveyor includes first and second tracks and wherein articles are disposed on each of the first and second tracks of the process conveyor at the same time that articles are disposed on the other one of the tracks of the process conveyor and wherein each of the first and second radiation sources radiates the articles on the first and second tracks of the process conveyor and wherein the articles are disposed on one of the tracks of the process conveyor in a first movement of the articles past one of the radiation sources when the other one of the radiation sources is not operative to irradiate the articles on the process conveyor and wherein xe2x80x83the articles are disposed on the other one of the tracks of the process conveyor in a second movement of the articles past the one of the radiation sources when the other one of the radiation sources is not operative to irradiate the articles on the process conveyor and wherein xe2x80x83the articles are transferred from the one of the tracks to the other one of the tracks and are inverted between the times of the first and second movements of the articles past the one of the radiation sources. 44. A method of irradiating articles including the steps of: providing a process conveyor, dividing the process conveyor into two (2) tracks, each movable at the same speed as the other track and each constructed to hold the articles to be irradiated, providing a radiation source constructed to irradiate articles, moving the process conveyor past the radiation from the radiation source to obtain an irradiation of the articles on the two tracks, and converging the articles on each of the tracks toward the divider during the movement of the process conveyor toward the radiation source to minimize the width of the radiation from the radiation source. 45. A method as set forth in claim 44 , including the steps of: claim 44 providing a load conveyor for conveying the articles toward the process conveyor and dividing the load conveyor into two (2) tracks corresponding to the two (2) tracks on the process conveyor, transferring the articles from the load conveyor to the process conveyor for the movement of the articles by the process conveyor toward the radiation source, diverging the articles on the two (2) tracks on the load conveyor during the movement of the articles on the load conveyor toward the process conveyor, and transferring the articles from a loading area to the load conveyor for movement of the articles to the process conveyor. 46. A method as set forth in claim 44 including the steps of: claim 44 providing a load conveyor for conveying the articles from the process conveyor after the irradiation of the articles by the radiation source, dividing the load conveyor into two (2) tracks corresponding to the two (2) tracks on the process conveyor, transferring the articles on the two (2) tracks on the process conveyor to the two (2) tracks on the load conveyor after the irradiation of the articles on the process conveyor, diverging the articles from the divider during the movement of the articles on the load conveyor, and transferring the articles from the level conveyor to an unloading area. 47. A method as set forth in claim 46 wherein claim 46 the process conveyor conveys the articles at a substantially constant speed past the radiation from the radiation source and wherein the load conveyors convey the articles at a different speed than the speed at which the articles are conveyed on the process conveyor. 48. A method as set forth in claim 44 wherein claim 44 the process conveyor is formed from three (3) segments disposed in a series relationship and wherein the second segment is separated by gaps from the first and third segments and wherein the gaps have a length less than the length of the articles and wherein first and second radiation sources are respectively disposed in the gaps separating the second segment from the first and third segments and wherein the first and second radiation sources are disposed on the opposite sides of the process conveyor and wherein xe2x80x83the first radiation source is pointed toward the article on the process conveyor in a direction opposite to the direction in which the second radiation source is pointed. 49. A method as set forth in claim 45 wherein claim 45 the articles are disposed in a non-uniform relationship to one another on the load conveyor and are transferred to the process conveyor for disposition on the process conveyor in a substantially uniform relationship to one another. 50. A method as set forth claim 45 wherein claim 45 rollers are provided on the process conveyor to converge the articles on the first and second tracks on the process conveyor toward the divider and wherein rollers are provided in the load conveyor to diverge the articles on the load conveyor from the divider on the load conveyor. 51. A method as set forth in claim 48 , including the steps of: claim 48 providing for one of the radiation sources to irradiate first and second opposite sides of the article when the other one of the first and second radiation sources is inoperative. 52. A method as set forth in claim 48 , including the steps of: claim 48 extending a third load conveyor from the first track of the second load conveyor to the second track of the first load conveyor, inverting the articles on the third load conveyor, and activating the third load conveyor to receive the articles on the first track of the second load conveyor and to invert the articles and to transfer the inverted articles to the second track on the first load conveyor to obtain a radiation of the second side of the articles on the second track of the first load conveyor. 53. A system for irradiating articles, including, a process conveyor for conveying the articles in a first direction and including a divider for separating the conveyor into first and second tracks, a source of radiation for providing radiation in a second direction transverse to the first direction, and structure for converging the articles in the two (2) tracks on the process conveyor toward the divider on the process conveyor during the movement of the articles on the process conveyor, thereby to limit the width of the radiation source. 54. A system as set forth in claim 53 wherein claim 53 the process conveyor is divided into three (3) segments disposed in a series relationship and wherein the second segment is respectively separated by first and second gaps from the first and third segments and wherein the gaps have a length less than the length of the articles and wherein the source of radiation comprises first and second sources of radiation and wherein xe2x80x83the first radiation source is disposed adjacent the first gap on a first side of the article and the second radiation source is disposed adjacent the second gap on a second side of the article opposite to the first side of the articles. 55. A system as set forth in claim 53 , including claim 53 a load conveyor disposed relative to the process conveyor for conveying the articles and for transferring the articles to the process conveyor, the load conveyor including a divider for separating the load conveyor into first and second tracks each constructed to convey articles at the same time as the conveyance of articles by the other track, the load conveyor being constructed to diverge the articles from the divider during the conveyance of the articles on the load conveyor. 56. A system as set forth in claim 53 , including claim 53 a load conveyor disposed relative to the process conveyor for conveying the articles after the irradiation of the articles by the source of radiation, the load conveyor including a divider for separating the load conveyor into first and second tracks each constructed to convey articles at the same time as the conveyance of articles by the other track, the load conveyor being constructed to diverge the articles from the dividers during the conveyance of the articles on the load conveyor. 57. A system as set forth in claim 55 , including claim 55 a loader for providing articles to the load conveyor, the process conveyor being constructed to convey the articles on the process conveyor at a substantially constant speed, the load conveyor and the process conveyor being disposed relative to each other and constructed to provide the conveyance of the articles on the load conveyor at a different speed than the speed of the articles on the process conveyor and to provide for a substantially uniform spacing of the articles on the process conveyor. 58. A system as set forth in claim 55 wherein claim 55 the articles are disposed in a non-uniform relationship to one another on the load conveyor and wherein a controller is provided in association with the load conveyor and the process conveyor to transfer the articles from the load conveyor to the process conveyor in a substantially uniform relationship to one another on the process conveyor. 59. A system as set forth in claim 55 wherein claim 55 rollers are provided on the process conveyor to converge the articles on the first and second tracks on the process conveyor toward the divider and wherein rollers are provided on the load conveyor to diverge the articles on the conveyor from the divider. 60. A system as set forth in claim 55 wherein claim 55 a second load conveyor is disposed relative to the process conveyor to receive the articles from the process conveyor after the irradiation of the articles and is provided with a divider to separate the second load conveyor into first and second tracks and wherein the articles are irradiated by a first radiation source disposed on a first side of the articles and by a second radiation source disposed on a second side of the articles opposite to the first side and wherein the first and second load conveyors are interrelated to provide for one of the radiation sources to irradiate the two (2) opposite sides of the articles when the other one of the radiation sources is inoperative. 61. A system as set forth in claim 60 , wherein claim 60 the interrelationship between the first and second load conveyors is provided by a third load conveyor to transfer the articles from the first track on the second load conveyor to the second track on the first load conveyor when the second radiation source is inoperative and wherein an inverter is disposed relative to the third load conveyor to invert the articles on the third load conveyor, thereby providing for an irradiation of the first side of the articles by the first radiation source during the disposition of the articles on the first track of the first load conveyor and for an irradiation of the second side of the articles by the second radiation source during the disposition of the articles on the second track of the first load conveyor. 62. A system for irradiating articles, including a source of radiation, a process conveyor for moving the articles on the process conveyor past the source of radiation for irradiating the articles, a divider on the process conveyor for dividing the process conveyor into first and second tracks, and members on the first and second tracks of the process conveyor for conveying the articles on the tracks toward the divider during the movement of the articles on the process conveyor toward the radiation from the source. 63. A system as set forth in claim 62 wherein claim 62 the radiation source is on a first side of the process conveyor to radiate a first side of the articles and wherein a second radiation source is on a second side of the process conveyor to radiate a second side of the articles opposite to the first side of the articles. 64. A system as set forth in claim 63 wherein claim 63 gaps are provided in the process conveyor at the positions of the radiation from the first and second sources. 65. A system as set forth in claim 62 wherein claim 62 the articles are moved by the process conveyor past the source of radiation at a substantially constant speed. 66. A system as set forth in claim 63 wherein claim 63 gaps are provided on the process conveyor at the positions of the radiation from the first and second sources, and the articles are moved by the process conveyor past the source of radiation at a substantially constant speed. 67. A method of irradiating articles, including the steps of: providing a divider on a process conveyor in a longitudinal direction corresponding to a direction of movement of the process conveyor, directing radiation toward the articles on the process conveyor during the movement of the articles on the process conveyor, and converging the articles on the process conveyor during their movement on the process conveyor toward the radiation from the source. 68. A method as set forth in claim 67 wherein claim 67 the process conveyor moves at a substantially constant speed past the radiation from the source. 69. A method as set forth in claim 67 wherein claim 67 the source of radiation constitutes a first source of radiation and wherein a second source of radiation is disposed on a second side of the articles opposite to the first side of the articles and wherein first and second gaps are provided in the process conveyor at spaced positions in the direction of movement of the process conveyor respectively corresponding to the position of the first and second sources of radiation. 70. A method as set forth in claim 69 wherein claim 69 the first and second sources of radiation are respectively disposed to pass radiation through the first and second gaps to the articles on the process conveyor and wherein the dimensions of the gaps in the direction of movement of the process conveyor is less than the dimension of the articles in the direction of movement of the process conveyor. 71. A method as set forth in claim 68 wherein claim 68 the source of radiation constitutes a first source of radiation and wherein a second source of radiation is disposed on a second side of the articles opposite to the first side of the articles and wherein first and second gaps are provided in the process conveyor at spaced positions in the direction of movement of the process conveyor, and wherein the first and second sources of radiation are respectively disposed to pass radiation through the first and second gaps and wherein the dimensions of the gaps in the direction of movement of the process conveyor is less than the dimension of the articles in the direction of movement of the process conveyor.
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044180365	summary	CROSS-REFERENCES TO RELATED APPLICATIONS This application is related to copending applications Ser. No. 217,060 entitled "Mechanical Spectral Shift Reactor" by W. J. Dollard et al.; Ser. No. 217,056 entitled "Latching Mechanism" by L. Veronesi; Ser. No. 217,056 entitled "Spectral Shift Reactor Control Method" by A. J. Impink, Jr.; Ser. No. 217,061 entitled "Spectral Shift Reactor" by W. R. Carlson et al.; Ser. No. 217,052 entitled "Displacer Rod For Use In A Mechanical Spectral Shift Reactor" by R. K. Gjertsen et al.; Ser. No. 217,053 entitled "Mechanical Spectral Shift Reactor" by D. G. Sherwood et al.; Ser. No. 217,275 entitled "Mechanical Spectral Shift Reactor" by J. F. Wilson et al.; Ser. No. 217,055 entitled "Hydraulic Drive Mechanism" by L. Veronesi et al.; and Ser. No. 217,059 entitled "Fuel Assembly For A Nuclear Reactor" by R. K. Gjertsen; all of which are filed Dec. 16, 1980 and to Ser. No. 228,007 entitled "Self-Rupturing Gas Moderator Rod For A Nuclear Reactor" by G. R. Marlatt, filed Jan. 23, 1981 all of which are assigned to the Westinghouse Electric Corporation. BACKGROUND OF THE INVENTION The invention relates to fuel assemblies for nuclear reactors and more particularly to fuel assemblies for use in spectral shift reactors. In typical nuclear reactors, reactivity control is accomplished by varying the amount of neutron absorbing material (poisons) in the reactor core. Generally, neutron absorbing control rods are utilized to perform this function by varying the number and location of the control rods with respect to the reactor core. In addition to control rods, burnable poisons and poisons dissolved in the reactor coolant can be used to control reactivity. In the conventional designs of pressurized water reactors, an excessive amount of reactivity is designed into the reactor core at start-up so that as the reactivity is depleted over the life of the core the excess reactivity may be employed to lengthen the core life. Since an excessive amount of reactivity is designed into the reactor core at the beginning of core life, neutron absorbing material such as soluble boron must be placed in excess reactivity. Over the core life, as reactivity is consumed, the neutron absorbing material is gradually removed from the reactor core so that the original excess reactivity may be used. While this arrangement provides one means of controlling a nuclear reactor over an extended core life, the neutron absorbing material used during core life absorbs neutrons and removes reactivity from the reactor core that could otherwise be used in a more productive manner such as in plutonium production. The consumption of reactivity in this manner without producing a useful product results in a less efficient depletion of uranium and greater fuel costs than could otherwise be achieved. Therefore, it would be advantageous to be able to extend the life of the reactor core without suppressing excess reactivity with neutron absorbing material thereby providing an extended core life with a significantly lower fuel cost. One such method of producing an extended core life while reducing the amount of neutron absorbing material in the reactor core is by the use of "Spectral Shift Control". As is well understood in the art, in one such method the reduction of excess reactivity (and thus neutron absorbing material) is achieved by replacing a large portion of the ordinary reactor coolant water with heavy water. This retards the chain reaction by shifting the neutron spectrum to higher energies and permits the reactor to operate at full power with reduced neutron absorbing material. This shift in the neutron spectrum to a "hardened" spectrum also causes more of the U.sup.238 to be converted to plutonium that is eventually used to produce heat. Thus, the shift from a "soft" to a "hard" spectrum results in more neutrons being consumed by U.sup.238 in a useful manner rather than by poisons. As reactivity is consumed, the heavy water is gradually replaced with ordinary water so that the reactor core reactivity is maintained at a proper level. By the end of core life, essentially all the heavy water has been replaced by ordinary water while the core reactivity has been maintained. Thus, the reactor can be controlled without the use of neutron absorbing material and without the use of excess reactivity at start-up which results in a significant uranium fuel cost savings. The additional plutonium production also reduces the U.sup.235 enrichment requirements. While the use of heavy water as a substitute for ordinary water can be used to effect the "spectral shift", the use of heavy water can be an expensive and complicated technology. While there exist in the prior art numerous ways of controlling a nuclear reactor, what is needed is a fuel assembly for use in controlling neutron moderation in a manner that provides for reduced uranium fuel costs and for an extended reactor core life. SUMMARY OF THE INVENTION A fuel assembly for a nuclear reactor comprises a 5.times.5 array of guide tubes in a generally 20.times.20 array of fuel elements. The guide tubes are arranged to accommodate either control rods or water displacer rods. The fuel assembly also comprises a plurality of Inconel and Zircaloy grids arranged to provide stability of the fuel elements and guide tubes while allowing the flow of reactor coolant therebetween.
	claims	1. A condensation chamber cooling system, comprising: a condensation chamber for a boiling water reactor, said condensation chamber having a maximum fill line and a minimum fill line; at least one heat exchanger disposed outside said condensation chamber; an elongated cooling module being disposed in said condensation chamber, said condensation chamber having an upper region with an evaporation space; said cooing module being disposed in said condensation chamber with said evaporation space being located above said maximum filling level of said condensation chamber; at least one riser pipe and one downpipe, each of said riser pipe and said downpipe having an upper end and a lower end and, each of said respective upper ends issuing into said evaporation space, and each of said respective lower ends extending in said condensation chamber below said minimum filling level; a first pressure line connecting said evaporation space to said heat exchanger; and a second pressure line connecting said heat exchanger to said condensation chamber; said condensation chamber; said pressure lines, said cooling module, and said heat exchanger forming a passive closed cooling circuit. 2. The condensation chamber cooling system according to claim 1, wherein said downpipe is nested within said riser pipe. 3. The condensation chamber cooling system according to claim 1, wherein said heat exchanger includes an evaporation condenser. 4. The condensation chamber cooling system according to claim 1, further comprising a blow-off line issuing below said minimum filling level in said condensation chamber. 5. The condensation chamber cooling system according to claim 1, further comprising a vacuum pump acting upon said cooling circuit. 6. The condensation chamber cooling system according to claim 1, wherein said at least one heat exchanger is disposed geodetically above said condensation chamber. 7. The condensation chamber cooling system according to claim 1, wherein said minimum filling level corresponds at least approximately to said maximum filling level. 8. The condensation chamber cooling system according to claim 1, further comprising an additional active cooling circuit connected to said condensation chamber, said additional active cooling circuit being provided for discharging waste heat from said condensation chamber. 9. The condensation chamber cooling system according to claim 1, further comprising a pumping system connected to said condensation chamber, said pumping system providing return of water located in said condensation chamber into the boiling water reactor. 10. The condensation chamber cooling system according to claim 1, wherein said heat exchanger is provided for discharging waste heat into a surrounding.
06132356&	abstract	The invention provides a portable apparatus and a method for quickly conting, suppressing and mitigating localized hazardous material spills. The hazardous material containment apparatus has a vapor, aerosol and particulate containment vessel having a cover and side walls attached around a perimeter of the cover to define an open central cavity. Each of the cover and side walls are composed of a material which resists penetration of a hazardous material therethrough. A tube extending through the vessel has a first end open at a point inside the cavity and a second end at a point outside the vessel. Vacuum and filtration systems are attached to the second end of the tube for extracting a hazardous material from inside the cavity to outside the vessel through the tube.
	description	This U.S. non-provisional patent application claims priority under 35 U.S.C. § 119 of Korean Patent Application Nos. 10-2016-0009297, filed on Jan. 26, 2016, and 10-2016-0088781, filed on Jul. 13, 2016, the entire contents of which are hereby incorporated by reference. The present disclosure relates to an X-ray tube, and more particularly, to an X-ray tube generating an X-ray, having a simple structure from which a element necessary for focusing an electron beam, such as a magnetic lens, is removed, and having a nanometer-scale focal spot. An X-ray source (a nano focus X-ray source) having a nanometer-scale focal spot is required for a non-destructive inspection of an object having a microstructure, such as a semiconductor chip. In general, a nano focus X-ray source includes an electron source (cathode) generating an electron beam, a focusing unit focusing the electron beam emitted from the electron source, and a target (anode) enabling the focused electron beam to collide with each other to generate an X-ray. Herein, since the electron beam travels inside the X-rays source which is in a vacuum state, a proper vacuum is maintained in a path from the X-rays source to the target by a vacuum container. The focusing unit is composed of a lens for focusing an electron beam, etc. Since an electrostatic lens has a limitation in demagnification due to an aberration etc., a magnetic lens having one or more stages for high focusing of electron beams is used to focus the electron beam in a nano meter size. The focused electron beam collides with a target of metal material and generates a nano focus X-ray. In general, a magnetic lens is bulky and heavy, and continuously consumes current in order to form a magnetic field. Thus, the related arts of using the magnetic lens as a focusing unit have a limitation in that an X-ray source has a bulky and heavy shape due to the magnetic lens. The present disclosure provides an X-ray tube having a simple structure from which an element required for focusing an electron beam, such as a lens is removed, and generating a nano focus X-ray. An embodiment of the inventive concept provides an X-ray tube including an electron beam generation unit emitting an electron beam, a limiting electrode unit limiting the electron beam emitted from the electron beam generation unit, and a target unit including a target material emitting an X-ray when the limited electron beam collides with the target material, wherein the limiting electrode unit includes an electron beam limiting electrode allowing a portion of the emitted electron beam to pass therethrough and to be delivered to the target unit. In an embodiment, the target unit may emit an X-ray having a focal spot corresponding to the size of the portion of the electron beam delivered to the target unit by the limiting electrode unit. In an embodiment, the limiting electrode unit may include a penetration type electron beam limiting electrode having a limiting opening having a predetermined diameter. In an embodiment, the penetration type electron beam limiting electrode may deliver, to the target unit, a portion of the electron beam having passed the limiting opening among the emitted electron beams. In an embodiment, the penetration type electron beam limiting electrode may be configured to have an equal electric potential as the target unit. In an embodiment, the limiting electrode unit may include at least one slit type electron beam limiting electrode having a slit having a predetermined width. In an embodiment, the at least one slit type electron beam limiting electrode may include at least one spacer having the thickness corresponding to the predetermined width, and a plurality of metal electrodes spaced by the at least one spacer. In an embodiment, the at least one slit type electron beam limiting electrode may be disposed such that the slits are aligned with each other at a predetermined angle. In an embodiment, the at least one slit type electron beam limiting electrode may have a matrix shape in which the slits are provided in a plurality of columns and rows. In an embodiment, the slit may have an incident surface into which the emitted electron beam is incident and an emitting surface through which the portion of the electron beam is delivered to the target unit, and a width of the incident surface is greater than a width of the emitting surface. In an embodiment, the limiting electrode unit may include at least one electron beam limiting electrode made of at least one of tungsten, molybdenum or gold. In an embodiment, the X-ray tube may further include an electrostatic polarizer disposed between the electron beam generation unit and the limiting electrode unit, and controlling an incident location of the emitted electron beam with respect to the limiting electrode. In an embodiment, the X-ray tube may further include a filter disposed between the target unit and an object to which the X-ray is delivered, and removing a low energy X-ray. In an embodiment, the filter may be integrally provided with the target unit. In an embodiment, the X-ray tube may further include a donut-shaped electrode disposed between the electron beam generation unit and the limiting electrode, and preventing retrogression of an X-ray by a remaining electron beam limited by the limiting electrode unit. In an embodiment, the electron beam generation unit may include a cathode emitting the electron beam, and the target unit comprises an anode. In an embodiment, the electron beam generation unit may further include a focusing unit focusing the emitted electron from the cathode in a micrometer-scale. In describing embodiments of the inventive concept, detailed descriptions related to well-known functions or configurations will be ruled out in order not to unnecessarily obscure subject matters of the inventive concept. Herein, the term “comprise”, “have”, “may comprise” or “may have” intends to mean that there may be specified features, numerals, steps, operations, elements, parts, or combinations thereof, not excluding the possibility of the presence or addition of the specified features, numerals, steps, operations, elements, parts, or combinations thereof. The term “include,” “comprise,” “may include,” or “may comprise” used herein indicates disclosed functions, operations, or existence of elements but does not limit one or more additional functions, operations or elements. Also, it should be further understood that the terms “include,” “comprise,” or “have” used herein are intended to specify the presence of stated features, integers, steps, operations, elements, components, and/or combinations thereof described in the specification, but are not intended to pre-exclude the presence or addition of one or more other features, integers, steps, operations, elements, components, and/or combinations thereof. Herein, a singular form, unless otherwise indicated in context, may include plural forms. Hereinafter, example embodiments of the inventive concept will be described in detail with reference to the accompanying drawings. FIG. 1 is a view showing a structure of a general X-ray tube. Referring to FIG. 1, a general X-ray tube 100 is configured to include an electron source 110, a focusing unit 120 focusing electron beams emitted from the electron source 110, and a target unit 130 consisting of a target material emitting an X-ray by collision of the electron beams focused in the focusing unit 120 thereto. The foregoing elements are installed in a vacuum container completely sealed or in which an internal vacuum state is maintained by a vacuum pump such that electron beams may be generated and accelerated in a vacuum environment. The vacuum container may be made of a ceramic or a glass material such as aluminum oxide, aluminum nitride or glass having excellent high voltage characteristics and suitable for a vacuum container. In general, the electron source 110 is composed of a cathode and is connected to a negative terminal of a power source, and the target unit 130 is composed of an anode and is connected to a positive terminal of the power source. Electrons emitted from the cathode are accelerated from the cathode to the anode by a difference between a negative potential of the cathode and a positive potential of the anode to form an electron beam 1. The electrons arriving at the anode collides with a metal target of the anode to generate an X-ray 3. In various related arts, a gate for controlling the amount of the electron beam may be further provided between the anode and the cathode. The focusing unit 120 includes at least one focusing lens to focus the electron beam 1 emitted from the electron source 110 in a required size. For example, the focusing unit 120 may include an electrostatic lens, a magnetic lens or the like. In general, the electrostatic lens has a limitation in reducing the size thereof, and the magnetic lens also has disadvantages of being bulky and heavy and continuously consuming current. In a non-destructive inspection of a fine structure of an object such as a semiconductor chip, it is required that the focusing unit 120 sufficiently focus the electron beam 1 in order to generate an X-ray (a nano focus X-ray) having a nanometer-scale focal spot. However, when a plurality of large sized focusing lenses are mounted in the focusing unit 120 for the foregoing purpose, the overall size of the X-ray tube 100 increases. Hereinafter, an X-ray tube structure capable of generating a nano focus X-ray without a focusing lens is described as a technical feature of the inventive concept to solve the foregoing limitations. In the followings embodiments, the present disclosure is characterized in that a focusing unit 120 is composed of an electron beam limiting electrode having a channel which allows only a portion of an electron beam 1 emitted from an electron source 110 to arrive at a target unit 130 and a remaining electron beam 1 to be blocked. FIG. 2 is a view showing a structure of an X-ray tube according to embodiment 1 of the inventive concept. Referring to FIG. 2, an X-ray tube 200 according to embodiment 1 of the inventive concept includes an electron beam generation unit 210 emitting an electron beam, a limiting electrode unit 220 limiting the electron beam emitted from the electron beam generation unit 210, and a target unit 230 composed of a target material emitting an X-ray by collision with the electron beam having passed through the limiting electrode 220 and having a limited size. The foregoing elements are installed in a vacuum container completely sealed or in which an internal vacuum state is continuously maintained by a vacuum pump such that electron beams may be generated and accelerated in a vacuum environment. The vacuum container may be made of a ceramic or a glass material, such as aluminum oxide, aluminum nitride or glass having excellent high voltage characteristics and suitable for a vacuum container. The electron beam generation unit 210 is composed of a cathode connected to a negative terminal of a power source and emitting an electron in the form of an electron beam. Also, the electron beam generation unit 210 may include a focusing unit configured to focus the electron beam emitted from the cathode such that the electron beam has a constant size, for example a nanometer-scale size. A gate configured to control the amount of the electron beam may be further provided to the electron beam generation unit 210. The target unit 230 is composed of an anode and is connected to a positive terminal of the power source. The electron beam generated from the electron beam generation unit 210 collides with a metal target of the anode to generate an X-ray. In the X-ray tube generating a nano focus X-ray, since the current of an electron beam arriving at the target unit 230 is several to several tens of micro amperes (μA) which are not relatively high, it is possible to generate a nano focus X-ray by limiting the diameter of the focused electron beam such that the focused electron beam has a nanometer-scale diameter as well as a sufficiently high current density. In embodiment 1 of the inventive concept, the limiting electrode unit 220 is composed of a penetration type electron beam limiting electrode 221 having a limiting opening 222 with a predetermined diameter. The diameter of the limiting opening 222 may be determined by the size of a focal spot of an X-ray to be generated in the X-ray tube 200. For example, the limiting opening 222 of embodiment 1 may have a nanometer-scale diameter, and in this case, the X-ray tube 200 may possibly generate a nano focus X-ray. In embodiment 1 of the inventive concept, the electron beam 1 emitted from the electron beam generation unit 210 may have a micrometer-scale diameter. The electron beam 1 emitted from the electron beam generation unit 210 is delivered to the limiting electrode unit 220. In the limiting electrode unit 220, only a portion of electron beam 2 having passed the limiting opening 222 arrives at the target unit 230, and a remaining electron beam is limited by the penetration type electron beam limiting electrode 221 and thus is unable to arrive at the target unit 230. Herein, when the diameter of the limiting opening 222 is nanometer-scale, the electron beam 2 arriving at the target unit 230 will have the nanometer-scale diameter, and consequently, an X-ray 3 having a nanometer-scale focal spot may be emitted from the target unit 230 by collision of the electron beam 2. When the electron beam limiting electrode 221 is a metal electrode type, the remaining electron beam limited by collision with the penetration type electron beam limiting electrode 221 may generate an unnecessary X-ray inside the X-ray tube 200. Thus, in various embodiments of the inventive concept, the penetration type electron beam limiting electrode 221 may be made of a material and with a thickness capable of shielding the X-ray generated by the limited rest electron beam. For example, the penetration type electron beam limiting electrode 221 may be made of a material having good shielding performance, such as tungsten, molybdenum or gold. Further, in various embodiments of the inventive concept, the penetration type electron beam limiting electrode 221 may be configured to have the same electric potential as the electric potential of the target unit 230. In order to produce the penetration type electron beam limiting electrode 221 according to embodiment 1 of the inventive concept, the limiting opening 222 having a predetermined diameter, such as a nanometer-scale diameter, should be precisely formed in a thick solid material capable of shielding an X-ray. Since the production process requires a high level of technology and a high accuracy, production efficiency may be lowered. Hereinafter, a nano focus X-ray structure according to another embodiment of the inventive concept to solve the limitation in production will be described. FIG. 3 is a view showing a structure of an X-ray tube according to embodiment 2 of the inventive concept. Referring to FIG. 3, an X-ray tube 300 according to embodiment 2 of the inventive concept includes an electron beam generation unit 310 emitting an electron beam, a limiting electrode unit 320 limiting the electron beam emitted from the electron beam generation unit 310, and a target unit 330 composed of a target material generating an X-ray by collision with the electron beam having passed the limiting electrode unit 320 and having a limited size. The foregoing elements are installed in a vacuum container completely sealed or in which an internal vacuum state is continuously maintained by a vacuum pump such that an electron beam may be generated and accelerated in a vacuum environment. The vacuum container may be made of a ceramic or a glass material, such as aluminum oxide, aluminum nitride or glass having an excellent high voltage characteristics and suitable for a vacuum container. The electron beam generation unit 310 is composed of a cathode connected to a negative terminal of a power source and emitting an electron in the form of an electron beam. Also, the electron beam generation unit 310 may include a focusing unit focusing the electron beam emitted from the cathode such that the electron beam has a constant size, for example a nanometer-scale size. A gate configured to control the amount of the electron beam may be further provided to the electron beam generation unit 310. The target unit 330 is composed of an anode and is connected to a positive terminal of the power source. The electron beam generated from the electron beam generation unit 310 collides with a metal target of the anode to generate an X-ray. In embodiment 2 of the inventive concept, the limiting electrode unit 320 is composed of a plurality of slit type electron beam limiting electrodes 321 and 322. The plurality of slit type electron beam limiting electrodes 321 and 322 each may include a slit having a predetermined width. For example, the plurality of slit type electron beam limiting electrodes 321 and 322 respectively include a plurality of metal electrodes 321b and 322b spaced by at least one spacer 321a and 322a having a predetermined thickness. The thickness of the at least one spacer 321a and 322a determines the slit width, and the slit width may be determined by the size of a focal spot of an X-ray to be generated by the X-ray tube 300. For example, in embodiment 2, the at least one spacer 321a and 322a may have the thickness of nanometer-scale, and in this case, it is possible that the X-ray tube 300 generates a nano focus X-ray. In the above, it is described as an example that the plurality of slit type electron beam limiting electrodes 321 and 322 include the at least one spacer 321a and 322a and an assembly of the plurality of metal electrodes 321b and 322b, but the inventive concept is not limited thereto. In various embodiments, the plurality of slit type electron beam limiting electrodes 321 and 322 may be manufactured as one element having a slit having a predetermined thickness. In various embodiments of the inventive concept, the plurality of slit type electron beam limiting electrodes 321 and 322 may be disposed such that slits formed in each of the slit type electron beam limiting electrodes 321 and 322 are aligned at an arbitrary angle with each other. For example, when the X-ray tube 300 is composed of two slit type electron beam limiting electrodes 321 and 322, the two slit type electron beam limiting electrodes 321 and 322 may be disposed such that the slits are aligned to be orthogonal to each other as illustrated in FIG. 3. In various embodiments, the alignment angle of the plurality of slits may be set to an angle at which a current amount of the electron beam 3 arriving at the target unit 330 is measured at a maximum level. In embodiment 2 of the inventive concept, the electron beam 1 emitted from the electron beam generation unit 310 may have a micrometer-scale diameter. The electron beam 1 emitted from the electron beam generation unit 310 is delivered to the limiting electrode unit 320. In the limiting electrode unit 320, only a portion of the electron beam 3 having passed the slit defined in the plurality of slit type electron beam limiting electrodes 321 and 322 arrives at the target unit 330 while the remaining electron beam is limited by the slit type electron beam limiting electrodes 321 and 322 and is unable to arrive at the target unit 330. In this case, when the slit width is nanometer-scale, the electron beam 3 arriving at the target unit 330 has a nanometer-scale diameter, and consequently, an X-ray 4 having a nanometer-scale focal spot may be emitted from the target unit 330 by collision of the electron beams 3. In an embodiment described with reference to FIG. 3, a portion of electron beam 2, among the electron beam 1 delivered to the limiting electrode unit 320, having passed through a slit in an x-axis direction defined in a first slit type electron beam limiting electrode 321 arrives at a second slit type electron beam limiting electrode 322. Also, a portion of the electron beam 3, among the electron beams 2 having arrived at the second slit type electron beam limiting electrode 322 and having passed through a slit in a y-axis direction defined in the second slit type electron beam limiting electrode 322 arrives at the target unit 330. As illustrated in FIG. 3, the portion of the electron beams 3 arriving at the target unit 330 after passing through all slits in the x-axis and y-axis directions has a shape having a nanometer-scale diameter. When an electron beam having high energy is focused during a long time on one position of the target unit 330 in the X-ray tube 300, the target material may be damaged. In order to prevent the target material from being damaged, in various embodiments of the inventive concept, the plurality of slit type electron beam limiting electrodes 321 and 322 may include slits provided in a matrix shape (a grid shape) of a plurality of columns and rows. In this case, the electron beam is selectively deflected and incident into one intersection among a plurality of intersections at which the plurality of columns and rows intersect, and in case that a spot of the target material at which the electron beam passing through the corresponding intersection arrives is damaged, the electron beam is moved to be deflected and incident into another intersection such that the electron beam arrives at a spot of another target material, thereby increasing life time of the target material. As in embodiment 1, the slit type electron beam limiting electrodes 321 and 322 in embodiment 2 may be made of a material and with a thickness capable of shielding an X-ray generated by the limited remaining electron beam. In various embodiments of the inventive concept, surfaces defining the slits of the plurality of metal electrodes 321b and 322b, i.e., surfaces facing to each other, may be machined to be smooth enough to define slits having a nanometer-scale width. Alternatively, spacers 321a and 322a determining the width of each slit may be precisely manufactured by a thin film forming method or a thick film forming method such as a chemical vapor deposition method or a physical vapor deposition method. In embodiment 2 of the inventive concept, the electron beam 1 having a micrometer-scale diameter incident into the plurality of slit type electron beam limiting electrodes 321 and 322 should be correctly incident into a location at which the slit is defined. Since an entrance in each of the slit type electron beam limiting electrodes 321 and 322 through which the electron beam 1 passes is very narrow slit type, it may be hard to locate the electron beam 1 correctly on the corresponding location. Also, since the amount of the electron beam 3 passing through the plurality of slits is less than the amount of the incident electron beam 1, the current of the electron beam 3 may not be sufficient to generate an X-ray 4. To solve such a limitation, in various embodiments of the inventive concept, an electrostatic polarizer having four or more phases may be provided between the plurality of electron beam limiting electrodes 321 and 322 and the electron beam generation unit 310 such that the location in the electron beam limiting electrodes 321 and 322 at which the electron beam emitted from the electron beam generation unit 310 is incident may be finely tuned. Alternatively, in various embodiments of the inventive concept, a slit type electron beam limiting electrode 420 may have a shape illustrated in FIG. 4. More particularly, the slit type electron beam limiting electrode 420 may have a shape in which having a slit width C in an incident surface of the electron beam is greater than a slit width D in an emitting surface of the electron beam. Herein, the slit width D in the emitting surface may be determined by the size of a focal spot of an X-ray to be generated in an X-ray tube, and for example, the slit width D in the emitting surface may be nanometer-scale. In this case, the total length B of the slit type electron beam limiting electrode 420 in a travelling direction of the electron beam, and the slit length A having the slit width D in the emitting surface may be formed at predetermined values so as to sufficiently shield a limited X-ray. In an embodiment described with reference to FIG. 4, a portion of the electron beam 1 emitted from the electron beam generation unit 310 may be lost by collision with an inner wall of the slit while travelling from the incident surface to the emitting surface. However, another portion of the electron beam collided with the inner wall of the slit is reflected, and finally passes through the slit. Therefore, the slit type electron beam limiting electrode 420 illustrated in FIG. 4 may enable a greater amount of electron beam to be emitted from the emitting surface than the slit type electron beam limiting electrodes 321 and 322 illustrated in FIG. 3. In this case, it is preferred that the diameter of the electron beam incident into the slit type electron beam limiting electrode 420 of FIG. 4 be formed to be less than the slit width C on the incident surface. In an embodiment described with reference to FIG. 4, most of electron beam incident with high energy may generate a large amount of heat by collision with the slit type electron beam limiting electrode 420. Since the generated heat may deform a physical shape of the slit type electron beam limiting electrode 420, the slit type electron beam limiting electrode 420 may be disposed outside the vacuum container as a protruding heat dissipation structure such that internal temperature of the X-ray tube 300 is prevented from rising and the heat is discharged to the outside. When an X-ray is used in semiconductor chip inspection equipment, a material, such as SiO2 used as an insulation film of a semiconductor, is deformed by absorbing an X-ray with low energy of several kV level, thereby causing a damage to the semiconductor chip. Since the low energy X-ray capable of damaging the semiconductor chip is highly scattered and may deteriorate an X-ray image quality in an inspection result, it is preferred to filter the low energy X-ray. In order to realize the above purpose, in various embodiments of the inventive concept, an X-ray tube may further include a SiO2 filter disposed between a target unit from which an X-ray is emitted and an object. In one embodiment, the SiO2 filter may be provided with the target unit. The SiO2 filter serves to eliminate an unnecessary low energy X-ray. Alternatively, in various embodiments of the inventive concept, the X-ray tube may have a target unit including an anode target formed on a SiO2 substrate. In this case, the SiO2 substrate may serve as an X-ray window while serving as an outer wall of a vacuum container of the X-ray tube. In various embodiments of the inventive concept, the X-ray tube may further include a donut-shaped electrode between an electron beam generation unit and a limiting electrode unit. In this case, an electron beam emitted from the electron beam generation unit passes through an internal hole of the donut-shaped electrode and arrives at the limiting electrode. The donut-shaped electrode may function to shield a backward retrogression of an X-ray generated by a limited electron beam by the limiting electrode unit. In embodiments set forth herein, it is described as an example that the target units 230 and 330 are provided as transmission-type targets, but the target units are not limited thereto. Embodiments of the inventive concept may be applicable to a case that the target units 230 and 330 are provided as reflection-type targets. The present disclosure enables to realize a nano focus X-ray tube with a simple structure without employing a complicated element such as a lens. While the present invention has been particularly shown and described with reference to exemplary embodiments thereof, it will be understood by those of ordinary skilled in the art that various changes may be made therein without departing from the scope of the present invention as defined by the following claims. Therefore, technical scope of the present invention should not be construed as limited to those described in the description, but determined by the appended claims.
	claims	1. A Thorium fuel rod assembly comprising:first and second support elements; anda plurality of Thorium fuel rods positioned between the first and second support elements, each Thorium fuel rod comprising:an outer fuel element rod containing a solid Thorium-containing material wherein the outer fuel element:has a longitudinal length;defines an interior cavity extending along at least a majority of the longitudinal length of the outer fuel element rod; anddefines a plurality of fins that project radially outwardly; andan inner core element formed from a Beryllium-containing material positioned within the interior cavity defined by the outer fuel element, wherein the inner core element:is generally tubular and has a longitudinal length greater than the longitudinal length of the outer fuel element rod such that at least a portion of the inner core element extends beyond a first end of the outer fuel element;defines an inner cavity extending along at least a majority of the longitudinal length of the inner core element;wherein the inner cavity of the Beryllium inner core element is configured to receive a beam of high energy particles and produce a (p,n) reaction resulting in the emission of a neutron; andwherein the outer fuel element rod is configured to receive the emitted neutron and cause fission of a Thorium nucleus in the outer fuel element rod. 2. The Thorium fuel rod assembly of claim 1 wherein the solid Thorium-containing material comprises metallic Thorium. 3. The Thorium fuel rod assembly of claim 1 wherein at least a plurality of the fins projecting radially from the plurality of fuel rods are configured to form a helical outer structure. 4. The Thorium fuel rod assembly of claim 1 wherein the inner core element extends along at least 75% of the length of the outer fuel element. 5. The Thorium fuel rod assembly of claim 1 wherein a shape of the inner cavity at any given radial cross section of the inner core element is a four-leaf clover. 6. The Thorium fuel rod assembly of claim 1 wherein wherein an inner surface of the inner core element includes spiral Beryllium projections along the longitudinal length of the inner core element. 7. The Thorium fuel rod assembly of claim 1 wherein the Thorium fuel rod assembly comprises at least five Thorium fuel rods, and wherein the at least five Thorium fuel rods are positioned between the first and second support elements such that one of the Thorium fuel rods is centrally positioned and the other four Thorium fuel rods are each located at the same radial distance from the centrally positioned Thorium fuel rod.
	abstract	In a method of constructing a head shield for a radiation machine, the angular distribution of radiation propagating from a source and the angular function of thickness of a material in attenuating the radiation to a certain level of its original value are determined. Based on the angular distribution of radiation from the source and the angular function of thickness of the material, the thicknesses of the material at a plurality of angular locations around the source and distances from the source can be calculated for attenuating the radiation to or less than a predetermined threshold value. A shield around the source is constructed based on the calculated thicknesses of the material through iterative steps to ensure a cost-saving, weight-efficient, optimal solution. A method of designing a local radiation shield for a point of interest in a radiation system is also described to improve the machine reliability, regardless of the motion component of the POI with respect to the main radiation source and the secondary radiation source created by the patient scatter.
	description	The invention relates to the field of power generation, including power module structures and support systems. In nuclear reactors designed with passive operating systems, the laws of physics are employed to ensure that safe operation of the nuclear reactor is maintained during normal operation or even in an emergency condition without operator intervention or supervision, at least for some predefined period of time. A nuclear reactor 5 includes a reactor core 6 surrounded by a reactor vessel 2. Water 10 in the reactor vessel 2 surrounds the reactor core 6. The reactor core 6 is further located in a shroud 122 which surround the reactor core 6 about its sides. When the water 10 is heated by the reactor core 6 as a result of fission events, the water 10 is directed from the shroud 122 and out of a riser 124. This results in further water 10 being drawn into and heated by the reactor core 6 which draws yet more water 10 into the shroud 122. The water 10 that emerges from the riser 124 is cooled down and directed towards the annulus 123 and then returns to the bottom of the reactor vessel 2 through natural circulation. Pressurized steam 11 is produced in the reactor vessel 2 as the water 10 is heated. A heat exchanger 135 circulates feedwater and steam in a secondary cooling system 130 in order to generate electricity with a turbine 132 and generator 134. The feedwater passes through the heat exchanger 135 and becomes super heated steam. The secondary cooling system 130 includes a condenser 136 and feedwater pump 138. The steam and feedwater in the secondary cooling system 130 are isolated from the water 10 in the reactor vessel 2, such that they are not allowed to mix or come into direct contact with each other. The secondary cooling system 130 may comprise piping 139 for transporting steam or feedwater. The reactor vessel 2 is surrounded by a containment vessel 4. The containment vessel 4 is designed so that water or steam from the reactor vessel 2 is not allowed to escape into the surrounding environment. A steam valve 8 is provided to vent steam 11 from the reactor vessel 2 into an upper half 14 of the containment vessel 4. A submerged blowdown valve 18 is provided to release the water 10 into suppression pool 12 containing sub-cooled water. Piping 139 and other connections are provided between the nuclear reactor 5 and the secondary cooling system 130 or other systems in the power generation facility. In the event of an earthquake or other seismic activity, significant forces or vibration may be transferred to, or by, the connections, which can place great stress on the connections. Forces resulting from thermal expansion also place stress on the connections. Maintaining integrity of these connections helps discourage the inadvertent release of radioactive or other materials from the various systems, and reduces maintenance or damage that might otherwise occur if one or more of the connections fail. The present invention addresses these and other problems. A power module is disclosed herein, as comprising a containment vessel completely submerged in a pool of liquid, and a support structure located at or above an approximate midpoint of the containment vessel, or center of gravity of the power module. The power module is supported by the support structure in combination with a buoyancy force of the pool of liquid acting on the containment vessel. A support structure for a containment vessel is disclosed herein, as comprising a support arm located at or above an approximate midpoint or center of gravity of the containment vessel, and a mounting structure submerged in water. The support structure further comprises a damping device disposed between the support arm and the mounting structure, wherein least a portion of the weight of the containment vessel is transferred to the mounting structure through the damping device. The damping device is configured to attenuate seismic forces transferred to the support arm. A system is disclosed herein, as comprising means for supporting a power module on a support structure, wherein the support structure is located at or above an approximate midpoint or center of gravity of the power module, and means for allowing a constrained rotation of the power module, wherein the support structure serves as a pivot for the rotation. The system further comprises means for damping seismic forces transmitted through the support structure to the power module. The invention will become more readily apparent from the following detailed description of a preferred embodiment of the invention which proceeds with reference to the accompanying drawings. Various embodiments disclosed or referred to herein may be operated consistent, or in conjunction, with features found in co-pending U.S. application Ser. No. 11/941,024 which is herein incorporated by reference in its entirety. FIG. 2 illustrates an example power module assembly comprising a containment vessel 24, reactor vessel 22 and a support structure 20. The containment vessel 24 is cylindrical in shape, and has ellipsoidal, domed or hemispherical upper and lower ends 26, 28. The entire power module assembly 25 may be submerged in a pool of liquid 36 (for example, water) which serves as an effective heat sink. The pool of liquid 36 is retained in reactor bay 27. The reactor bay 27 may be comprised of reinforced concrete or other conventional materials. The pool of liquid 36 and the containment vessel 24 may further be located below ground 9. The upper end 26 of the containment vessel 24 may be located completely below the surface of the pool of liquid 36. The containment vessel 24 may be welded or otherwise sealed to the environment, such that liquids and gas do not escape from, or enter, the power module assembly 25. The containment vessel 24 is shown suspended in the pool of liquid 36 by one or more support structures 20, above a lower surface of the reactor bay 27. The containment vessel 24 may be made of stainless steel or carbon steel, and may include cladding. The power module assembly 25 may be sized so that it can be transported on a rail car. For example, the containment vessel 24 may be constructed to be approximately 4.3 meters in diameter and 17.7 meters in height (length). Refueling of the reactor core 6 (FIG. 1) may be performed by transporting the entire power module assembly 25 by rail car or overseas, for example, and replacing it with a new or refurbished power module assembly which has a fresh supply of fuel rods. The containment vessel 24 encapsulates and, in some conditions, cools the reactor core 6 (FIG. 1). The containment vessel 24 is relatively small, has a high strength and may be capable of withstanding six or seven times the pressure of conventional containment designs in part due to its smaller overall dimensions. Given a break in the primary cooling system of the power module assembly 25 no fission products are released into the environment. The power module assembly 25 and containment vessel 24 are illustrated as being completely submerged in the pool of liquid 36. All sides, including the top and bottom, of the containment vessel 24 are shown as being in contact with, and surrounded by, the liquid 36. The one or more support structures 20 are located at an approximate midpoint of the containment vessel 24. In one embodiment, the one or more support structures 20 are located at an approximate center of gravity (CG), or slightly above the CG, of the power module 25. The power module 25 is supported by the support structure 20 in combination with a buoyancy force of the pool of liquid 36 acting on the containment vessel 24. In one embodiment, the power module assembly 25 is supported by two support structures 20; the first support structure located on a side of the power module assembly 25 opposite the second support structure. The one or more support structures 20 may be configured to support both the containment vessel 24 and the reactor vessel 22. In one embodiment, the one or more support structures 20 are located at an approximate CG, or slightly above the CG, of the reactor vessel 22. FIG. 3 illustrates a side view of the power module assembly 25 of FIG. 2. The containment vessel 24 as well as the reactor vessel 22, may be configured to pivot about the support structure 20, due to a rotational force RF acting on the power module 25. In one embodiment, the support structure 20 is located slightly above the CG of the power module 25, so that the lower end 28 tends to return to a bottom facing position within the reactor bay 27 due to gravity after the rotational force RF has subsided. The rotation of the containment vessel 24 also allows for greater maneuverability during installation or removal of the power module assembly 25 from the reactor bay 27. In one embodiment, the containment vessel 24 may be rotated between a vertical and a horizontal orientation or position of the power module assembly 25. The power module 25 is further illustrated as comprising a base skirt 30 located at the lower end 28 of the containment vessel 24. The base skirt 30 may be rigidly mounted to, welded on, or be an integral part of, the containment vessel 24. In one embodiment, the base skirt 30 is designed to support the weight of the power module 25 if the base skirt 30 is placed on the ground, on a transport device, or in a refueling station, for example. During normal operation (e.g. power operation) of the power module 25, the base skirt 30 may be suspended off the ground or positioned above the bottom of the reactor bay 27, such that the base skirt 30 is not in contact with any exterior component or surface. When the power module 25 rotates about the support structure 20, the lower end 28 of the containment vessel 25 tends to move in a lateral or transverse direction Lo. The base skirt 30 is configured to contact an alignment device 35 located in the pool of liquid 36 when the containment vessel 24 pivots a predetermined amount about the support structure 20. For example, the alignment device 35 may be sized so that the power module 25 is free to rotate within a range of motion or particular angle of rotation. The alignment device 35 may comprise an exterior diameter that is smaller than an interior diameter of the base skirt 30. The alignment device 35 may be sized to fit within the base skirt 30, such that the base skirt 30 does not contact the alignment device 35 when the power module 25 is at rest. In one embodiment, the base skirt 30 contacts the alignment device 35 when the containment vessel 24 pivots about the support structure 20. The base skirt 30 may not inhibit a vertical range of motion of the containment vessel 23, in the event that a vertical force acts upon the power module 25. The alignment device 35 may be rigidly mounted (e.g. bolted, welded or otherwise attached) to the bottom of the reactor bay 27. In one embodiment, one or more dampeners 38 are located between the base skirt 30 and the alignment device 35 to attenuate a contact force between the base skirt 30 and the alignment device 35 when the power module 25 pivots or rotates. The one or more dampeners 38 may be mounted to or otherwise attached to either the alignment device 35 (as illustrated) or the base skirt 30. FIG. 4 illustrates a partial view of an example support structure 40 for a power module assembly comprising a seismically isolated containment vessel 24. The support structure 40 comprises a support arm 45 and a mounting structure 47. The support arm 45 may be located at an approximate midpoint of the containment vessel 24. The mounting structure 47 is submerged in liquid (for example water), wherein the liquid surrounds the containment vessel 24. The mounting structure 47 may be an extension of, mounted to, recessed in, or integral with, the wall of the reactor bay 27 (FIG. 2). A damping device 46 is disposed between the support arm 45 and the mounting structure 47. At least a portion of the weight of the containment vessel 24 is transferred to the support structure 47 through the damping device 46. Damping device 46 may be elastic, resilient or deformable, and may comprise a spring, pneumatic or hydraulic shock absorber, or other vibration or force attenuating device known in the art. In one embodiment, the damping device 46 comprises natural or synthetic rubber. The damping device 46 may comprise an elastic material that is manufactured from petroleum or other chemical compounds and that is resistant to material breakdown when exposed to radiation or humidity. In yet another embodiment, the damping device 46 comprises soft deformable metal or corrugated metal. The damping device 46 is configured to attenuate dynamic or seismic forces transferred by and between the support arm 45 and the mounting structure 47. For example, a vertical or longitudinal force FV, acting along a longitudinal or lengthwise direction of the containment vessel 24, may act through the damping device 46. Additionally, a horizontal or transverse force FH may be exerted on the damping device 46 in any direction perpendicular to the longitudinal force FV. Transverse force FH may be understood to include a direction vector located in the plane defined by the X and Z coordinates of illustrative coordinate system 48, whereas the longitudinal force FV may be understood to include a direction vector oriented in the Y coordinate, the Y coordinate being perpendicular to the X-Z plane of the illustrative coordinate system 48. In one embodiment, by placing the support arm 45 at an approximate center of gravity of the containment vessel 24, a transverse force FH acting on the power module 25 tends to cause the containment vessel 24 to slide rather than rotate. Locating the support arm 45 on the containment vessel 24 at a particular height or position provides for controllability for how the containment vessel 24 will behave when it is subjected to one or more forces FH, FV, or RF. The damping device 46 may compress in a vertical direction to absorb or attenuate the longitudinal force FV. In one embodiment, the damping device 46 compresses or flexes in a horizontal direction to attenuate the transverse force FH. The damping device 46 may be configured to slide along the mounting structure 47 within the X-Z plane during a seismic activity, such as an earthquake or explosion. Forces FV and FH may also be understood to result from thermal expansion of one or more components of the power module 25, including containment vessel 24 (FIG. 2), in any or all of the three dimensions X, Y, Z. As a result of the compression or movement of the damping device 46, less of the forces FV and FH are transferred from the mounting structure 47 to the containment vessel 24, or from the containment vessel 24 to the mounting structure 47. The containment vessel 24 experiences less severe shock than what might otherwise be transferred if the support arm 45 were rigidly mounted to, or in direct contact with, the mounting structure 47. The containment vessel 24 may be configured to rotate about the horizontal axis X, due to a rotational force RF acting on the power module 25 (FIG. 3). Support arm 45 may be rigidly attached to the containment vessel 24, wherein the one or more elastic damping devices 46 is located between, and in contact with, both the support arm 45 and the mounting structure 47 located in the liquid 36 (FIG. 2). The elastic damping device 46 may provide a pivot point between the support arm 45 and the support structure 47, wherein the containment vessel 24 pivots or rotates about the elastic damping device 46, similar to that illustrated by FIG. 3. The weight of the containment vessel 24 may be supported, in part, by a buoyancy force of the liquid 36. The surrounding liquid 36 (FIG. 2) also serves to attenuate any of the transverse force FH, longitudinal force FV, and rotational force RF acting on the containment vessel 24. In one embodiment, the support arm 45 comprises a hollow shaft 29. The hollow shaft may be configured to provide a through-passage for an auxiliary or secondary cooling system. For example, piping 139 of FIG. 1 may exit the containment vessel 24 via the hollow shaft 29. FIG. 5 illustrates a partial view of a support structure 50 for a seismically isolated containment vessel 24 comprising a support arm 55 and multiple elastic damping devices 52, 54. The first elastic damping device 52 is located between the support arm 55 and a lower mounting structure 57. The second elastic damping device 54 is located between the support arm 55 and an upper mounting structure 58. In one embodiment, the first and second elastic damping devices 52, 54 are mounted to or otherwise attached to the support arm 55. In another embodiment, one or both of the first and second elastic damping devices 52, 54 are mounted to the lower and upper mounting structures 57, 58, respectively. At least a portion of the weight of the containment vessel 24 is transferred to the lower support structure 57 through the first elastic damping device 52. The first elastic damping device 52 may be under compression when the containment vessel 24 is at rest. The first elastic damping device 52 may be understood to attenuate longitudinal force acting between the support arm 55 and the lower mounting structure 57. The second elastic damping device 52 may also be understood to attenuate longitudinal force acting between the support arm 55 and the upper mounting structure 58. A longitudinal or vertical movement of the containment vessel 24 may be constrained by the lower and upper mounting structures 57, 58 as they come into contact with, or cause a compression of, the first and second elastic damping devices 52, 54, respectively. First and second elastic damping devices 52, 54 may provide similar functionality as a snubber or pair of snubbers in a conventional shock absorber. In one embodiment, the lower mounting structure 57 comprises a recess 56. The recess 56 may be sized such that it has an interior dimension or diameter that is larger than an exterior dimension or diameter of the first elastic damping device 52. The first elastic damping device 52 is illustrated as being seated or located in the recess 56. The recess 56 may operate to constrain a movement of the containment vessel 24 in one or more lateral or transverse directions. The first elastic damping device 52 may compress or flex when it presses up against a wall of the recess 56. In one embodiment, the recess 56 may restrict an amount or distance that the first elastic damping device 52 is allowed to slide on the lower mounting structure 57 when the containment vessel 24 experiences lateral or transverse force. FIG. 6 illustrates a partial view of an elastic damping and retaining structure 60 for a seismically isolated containment vessel 24. The damping and retaining structure 60 comprises a deformable portion 66. The deformable portion 66 may be dome shaped, elliptical or hemispherical in shape. Mounting structure 67 comprises a recess 68, wherein the deformable portion 66 is seated or located in the recess 68. The deformable portion 66 and recess 68 may be understood to function similarly as a ball joint, wherein the deformable portion 66 rotates or pivots within the recess 68. The recess 68 is illustrated as being concave in shape. The mounting structure 67 is configured to constrain a movement of the containment vessel 24 as a result of transverse force FH being applied in a lateral plane identified as the X-Z plane in the illustrative coordinate system 48. Additionally, the mounting structure 67 is configured to constrain a longitudinal movement of the containment vessel 24 as a result of a longitudinal force FV being applied in a direction Y perpendicular to the X-Z plane. The containment vessel 24 may be configured to rotate about the horizontal axis X, due to a rotational force RF acting on the power module 25 (FIG. 3). In one embodiment, the recess 68 forms a hemispherical, domed or elliptical bowl. A base skirt 30 (FIG. 2) located at the bottom end 28 of the containment vessel 24 may be configured to constrains a rotation of the containment vessel 24 as the deformable portion 66 pivots or rotates in the recess 68. The mounting structure 67 may be configured to support some or all of the weight of the power module 25 (FIG. 2). In one embodiment, a buoyancy force of the liquid 36 supports substantially all of the weight of the power module 25, such that the recess 68 of the mounting structure 67 may primarily operate to center or maintain a desired position of the power module 25. FIG. 6A illustrates a partial view of the elastic damping and retaining structure 60 of FIG. 6 responsive to a longitudinal force FV. The recess 68 in the mounting structure 67 comprises a radius of curvature R2 that is greater than a radius of curvature R1 of the deformable portion 66 of the damping and retaining structure 60 when the containment vessel 24 (FIG. 6) is at rest. Longitudinal force FV may be applied to the support arm 65 (FIG. 6) as a result of vertical movement of the containment vessel 24, or as a result of force transmitted from the mounting structure 67 to the containment vessel 24. The longitudinal force may result from an earthquake or explosion for example. When a dynamic longitudinal force FV is applied to the support arm 65, the damping device compresses from a static condition illustrated in solid lines by reference number 66, to a dynamic condition illustrated in dashed lines by reference number 66A. The radius of curvature of the deformable portion 66 temporarily approximates the radius of curvature R2 of the recess 68 in the dynamic condition 66A. As the effective radius of the deformable portion 66 increases, this results in an increased contact surface to form between the deformable portion 66 and the recess 68. As the contact surface increases, this acts to resist or decrease additional compression of the deformable hemispherical portion 66, and attenuates the longitudinal force FV. In one embodiment, the effective radius of curvature of the deformable hemispherical portion 66 increases with an increase in longitudinal force FV. When the dynamic longitudinal force FV has attenuated, the deformable portion 66 retains its original radius of curvature R1. FIG. 6B illustrates a partial view of the elastic damping and retaining structure 60 of FIG. 6 responsive to a transverse force FH. The recess 68 constrains a movement of the deformable portion 66 in at least two degrees of freedom. For example, the movement of the deformable portion 66 may be constrained in the X and Z directions of the illustrative coordinate system 48 of FIG. 6. The deformable portion 66 may compress or flex when it presses up against a wall of the recess 68. The compression or deformation of the deformable portion 66 attenuates the horizontal force FH. In one embodiment, the recess 68 may restrict an amount or distance that the deformable portion 66 is allowed to slide on the mounting structure 67 when the containment vessel 24 experiences transverse force FH. When a transverse force FH is applied to the support arm 65, the damping device moves or slides from the static condition illustrated in solid lines by reference number 66, to the dynamic condition illustrated in dashed lines by reference number 66B. Whereas the recess 56, 68 are illustrated in FIGS. 5 and 6 as being formed in the mounting structure 57, 67, other embodiments include where the recess 56, 68 is formed in the support arm 55, 65, and wherein the damping device 52, 66 is mounted to the mounting structure 57, 67. These alternate embodiments may otherwise operate similarly as the embodiments illustrated in FIG. 5 or 6, to constrain movement of the containment vessel 24 in one or both of the transverse and longitudinal directions. FIG. 7 illustrates a partial view of an elastic damping and retaining structure 70 for a seismically isolated power module 80. The power module 80 comprises a reactor vessel 22 and a containment vessel 24. The elastic damping and retaining structure 70 comprises one or more support arms, or trunnions, and one or more mounting structures. A first trunnion 75, protrudes or extends from the reactor vessel 22. The reactor vessel trunnion 75 provides similar functionality as one or more of the support arms described above with respect to FIGS. 2-6. A second trunnion 85 protrudes or extends from the containment vessel 24. The reactor vessel trunnion 75 lies along the same, single axis of rotation as the containment vessel trunnion 85. The axis of rotation X is shown in illustrative coordinate system 48. One or both of the reactor vessel 22 and containment vessel 24 may rotate about the axis of rotation X when a rotational force RF acts on the power module 25. The reactor vessel 22 and containment vessel 24 may rotate in the same or in opposite rotational directions from each other. Reactor vessel trunnion 75 is shown supported on a first mounting structure 77. The mounting structure 77 protrudes or extends from the containment vessel 24. The reactor vessel trunnion 75 may move or slide along the mounting structure 77 when horizontal force FH1 or FH2 acts on the power module 80. A first damping element 76 acts to attenuate or reduce the impact of horizontal force FH2 transmitted by or between the reactor vessel 22 and containment vessel 24. The first damping element 76 also helps to center or maintain a respective position or distance between the reactor vessel 22 and containment vessel 24 when the power module 80 is at rest or in a static condition. Containment vessel trunnion 85 is shown supported on a second mounting structure 87. In one embodiment, the mounting structure 87 protrudes or extends from a reactor bay wall 27. The containment vessel trunnion 85 may move or slide along the mounting structure 87 when horizontal force FH1 or FH2 acts on the power module 80. A second damping element 86 acts to attenuate or reduce the impact of horizontal force FH1 transmitted by or between the containment vessel 24 and the reactor bay wall 27. The second damping element 86 also helps to center or maintain a respective position or distance between the containment vessel 24 and the reactor bay wall 27 when the power module 80 is at rest or in a static condition. The first damping element 76 is shown housed in the reactor vessel trunnion 75. A reactor vessel retaining pin 90 is located in the reactor vessel trunnion 75 to provide a contact surface for the first damping element 76. The reactor vessel retaining pin 90 may be an extension of the containment vessel 24 or the containment vessel trunnion 85, for example. In one embodiment, the reactor vessel retaining pin 90 is rigidly connected to the containment vessel 24. The reactor vessel retaining pin 90 may extend through both sides of the containment vessel 24. Horizontal force FH2 may be transmitted by or between the reactor vessel 22 and the containment vessel 24 via the reactor vessel retaining pin 90 and the first damping element 76. Vertical movement of the reactor vessel 22 and containment vessel may be constrained by the interaction between the reactor vessel trunnion 75, reactor vessel retaining pin 90, and the mounting structure 77. Vertical movement of the reactor vessel 22 and containment vessel 24 may be further constrained by the interaction between the containment vessel trunnion 85 and the mounting structure 87. The elastic damping and retaining structure 70 may further operate to provide a thermal buffer for the power module 80. In addition to attenuating, damping, or otherwise reducing dynamic and seismic forces from being transferred to or between the components of the power module 80, the elastic damping and retaining structure 70 may reduce the thermal heat transfer between the reactor vessel 22 and the containment vessel 24. For example, one or both of the first and second mounting structures 77, 87 may be lined with thermal insulation. FIG. 8 illustrates a novel system 200 for seismically isolating a power module. The system 200 may be understood to operate with, but not limited by, means illustrated or described with respect to the various embodiments illustrated herein as FIGS. 1-7. At operation 210, a power module is supported on a support structure. The support structure may be located at or slightly above an approximate midpoint, or an approximate center of gravity, of the power module. At operation 220, rotation of the power module is constrained. The support structure may serve as a pivot for the rotation. At operation 230, seismic forces transmitted through the support structure to the power module are damped or attenuated. In one embodiment, the seismic forces are attenuated by a damping device comprising an elastic material. At operation 240, movement of the power module in one or more transverse directions is constrained within a fixed range of motion. Upon an attenuation of a transverse force, the power module returns to its original at-rest position. In one embodiment, the damping device comprises a rounded surface, and the support structure comprises a rounded recess configured to house the rounded surface. At operation 250, movement of the power module in a longitudinal direction is constrained within a fixed range of motion. Upon an attenuation of a longitudinal force, the power module returns to its original at-rest position. The longitudinal directional is perpendicular to the one or more transverse directions of operation 240. Although the embodiments provided herein have primarily described a nuclear reactor, it should be apparent to one skilled in the art that the embodiments may be applied to other types of power systems as described or with some obvious modification. Dimensions of the figures are not provided to scale, and in some cases certain features have been exaggerated in scale in order to illustrate or describe certain details. Other rates and values may be determined through experimentation such as by construction of full scale or scaled models of a nuclear reactor. Having described and illustrated the principles of the invention in a preferred embodiment thereof, it should be apparent that the invention may be modified in arrangement and detail without departing from such principles. We claim all modifications and variation coming within the spirit and scope of the following claims.
H00009369	summary	BACKGROUND OF THE INVENTION The present invention relates generally to a method for generating electricity from a fusion reactor and more particularly to a method for direct conversion of alpha-particle energy into electricity in a stellarator reactor. The guest to tap the energy of nuclear fusion by employing magnetic fields to confine an ultrahot plasma and generating electric power has been in progress for more than three decades. Several toroidal magnetic confinement fusion devices have been proposed. One such device is the tokamak where a toroidal current induced inside the plasma both heats the plasma and provides the poloidal magnetic field. There are several drawbacks, however, associated with the tokamak. The large plasma current needed for confinement in a tokamak, carries a large free energy that can be tapped by instabilities which destroy the confinement. Another problem associated with the plasma current is that it must be maintained by some means other than a transformer, since otherwise the pulse length is limited by the number of volt-seconds in the transformer windings. Another toroidal confinement machine is the stellarator, where the poloidal field is produced externally to the plasma by current-carrying conductors wound helically around the torus. This configuration does not require the large plasma current needed in a tokamak. Stellarators are capable of achieving betas several times greater than betas achievable in a tokamak. Stellarators are also capable of steady-state operation. A major problem in the design of a commercial fusion stellarator reactor, as with other conventional power sources, is the conversion of the thermal energy produced into electrical energy. Conventional designs call for the use of the thermal energy produced by fusion reactions to convert water to steam. The steam is used in a dynamic conversion processes to drive turbines and turbogenerators. This dynamic conversion process requires turbines, pumps, generators, large cooling systems and extensive piping systems. This auxiliary equipment is expensive, unreliable and relatively inefficient. Direct energy conversion techniques for tokamaks have been suggested in Fusion Energy Conversion, by George H. Milley, published by the American Nuclear Society, 1976. The problems associated with tokamaks, however, have been discussed above. Therefore, in view of the above, it is an object of the present invention to provide a novel cycle of operation for a stellarator fusion reactor. It is another object of the present invention to provide a novel cycle of operation for a stellarator fusion reactor for directly converting fusion energy into electrical energy. It is another object of the present invention to provide a novel cycle of operation of a stellarator fusion reactor which may be used in advanced neutron-beam fueled reactors. Is is a further object of the present invention to provide a cycle of operating two stellarators in tandem, such that the cycle is self-sustaining. It is still another object of the present invention to provide a stellarator reactor capable of directly generating electricity. It is still a further object of the present invention to provide a stellarator reactor system which is self-sustaining. Additional objects, advantages and novel features of the invention 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 maybe realized and attained by means of the instrumentalities and combinations particularly pointed out in the appended claims. SUMMARY OF THE INVENTION To achieve the foregoing and other objects in accordance with the purpose of the present invention, as embodied and broadly described herein, the method of this invention may comprise a three step process in which the minor radius of a stellarator is compressed and expanded. In the first step an ignited plasma is in thermal balance. The plasma is compressed adiabatically and the plasma .beta. decreases. In the next step the plasma volume is kept constant and the plasma temperature and 62 are driven up by the excess thermonuclear alpha-particle heating. As .beta. approaches the maximum .beta. attainable, the rate of energy loss increases until it balances the alpha-particle heating power and the plasma is again in a state of thermal balance. In the final step the plasma is expanded back to its original radius. When the plasma expands the corresponding pressure is higher than the corresponding pressure during compression since .beta. stays at .beta..sub.c during the expansion. Therefore, negative work is done on the plasma over the complete cycle. This work manifests itself as a back-voltage in the toroidal field coils and direct electrical energy is obtained from this voltage. As an alternate cycle, net work can also be done on the external system by allowing the plasma to expand at a constant pressure in the second step of the method, rather than keeping the plasma at constant volume. A magnetic confinement fusion reactor for directly generating electricity using the methods of the present invention comprises: a vacuum vessel; helical stabilizing coils; toroidal confining coils; means for generating a current through the toroidal coils; means for compressing a plasma disposed in the vacuum vessel; means for maintaining the volume of the plasma constant; means for expanding the plasma and means for transmitting current generated by the plasma from the toroidal coils. By operating two or more reactors in tandem, such that part of the energy produced by one reactor is used to compress and/or maintain a constant plasma volume in a second reactor the cycle can be made self-sustaining. The present invention provides a method and apparatus for obtaining electrical energy directly from a stellerator fusion reactor. Therefore, the present invention obviates the need for any of the intervening machinery associated with a turbogenerator.
042648238	summary	BACKGROUND OF THE INVENTION This invention relates to pulsed neutron well logging and more particularly to means for controlling the neutron output of a neutron generator tube used in pulsed neutron well logging. In recent years pulsed neutron well logging has become a commercially important well logging technique. Pulsed neutron techniques have been utilized for measuring the thermal neutron lifetime or thermal neutron decay time of earth formations in the vicinity of a well borehole, for making activation analyses of elemental constituents of the earth formations in the vicinity of the well borehole, for making porosity measurements of the earth formations in the vicinity of the well borehole and for making inelastic neutron scattering measurements for fast neutrons. In each of these well logging techniques the pulsed neutron source used to generate neutron pulses for the physical measurements has typically been an evacuated tube, deuterium-tritium accelerator type source. Other techniques such as those disclosed in U.S. Pat. No. 3,940,611 call for waveforms other than square wave pulses to be produced by a neutron generator tube. The system of the present invention is capable of providing such other waveforms as may be desired. Such sealed off or evacuated tube neutron sources generally comprises an outer envelope of glass, metal or some other vacuum encapsulation material, such as ceramic, which houses therein the elements of the neutron generator tube. The elements generally comprise a target which is electrically insulated at a high voltage potential, a source of ions which may be accelerated onto the target by its high voltage potential and a pressure regulator or replenisher element which may be used to stabilize or control the amount of pressure of gas within the evacuated outer envelope. Gas pressures of about 10.sup.-2 mm Hg. are typical for the operation of these tubes. The replenisher or pressure regulator of neutron generator tubes generally comprises a heater element which is surrounded by a surface which is capable of absorbing or emitting gas molecules of the gas filling the evacuated tube envelope as a function of its temperature. The capability of such a surface for emitting or absorbing gases in the tube envelope is controlled by the temperature of a heating element associated with it. When the heating element is elevated in temperature, the surrounding gas impregnated surface is encouraged to dispel absorbed gases by thermal emission. When the heating element is cooled, the surrounding surfaces associated with it are encouraged to absorb gases from the atmosphere inside the evacuated tube envelope. The amount of gas present in the tube envelope controls the amount of gas present in the ion source and hence, the capability of the ion source to produce positively charged ions of gas for acceleration onto the target material. In a typical neutron generator tube operation, the gas present in the evacuated envelope may be either deuterium gas or a mixture of deuterium and tritium gas. The target material is impregnated with tritium. Thus when deuterium ions are formed in the ion source and accelerated onto the target by its high voltage potential, the electrostatic Coulomb repulsion between the ions being accelerated and the nuclei of the tritium atoms is overcome and nuclear fusion takes place. This produces the unstable isotope helium 5 which immediately decays by the emission of an approximately 14 MEV monoenergetic neutron characteristic of this decay. A problem which has been associated with the use of such neutron generator tubes in well logging has been that the output of the neutron generator falls off as a function of time as the tritium in the target material is effectively used by the nuclear reactions and by heating of the target. Also high voltage power supply voltage variations, replenisher current variations and ion source emission capability can cause neutron output to vary. For most well logging operations it is highly desirable that during a given logging run the average neutron output of the tube remain constant and also as high as possible. High output is desirable to promote the nuclear interactions sought to be measured by the well logging technique in use. Consistency of the neutron output is desirable to promote measurement consistency and to avoid systematic errors. BRIEF DESCRIPTION OF THE INVENTION The average neutron output of a neutron generator tube is a function of the average target current of the tube. The target current, in turn, is a function of the target high voltage, the ion source voltage and the replenisher heater current. In the present invention, the target high voltage is at a fixed value. The ion source voltage is varied in a preprogrammed manner to produce neutron pulses or other waveforms. By varying the replenisher heater current and hence its heater element temperature, the average neutron output of the tube is controlled. In a preferred embodiment of the present invention, the target beam current is monitored, converted to a voltage signal and compared to a preprogrammed reference voltage control function. An error voltage developed from this comparison is used to control the replenisher current in such a manner that the replenisher current is automatically adjusted to maintain a constant value of the average target beam current corresponding to the reference voltage. A circuit for accomplishing this is provided which may be described as a series regulation replenisher current control circuit. Such a circuit may be used to vary the replenisher heater current or even turn off the replenisher current completely. The circuit embodiments of the present invention provide advantages over prior art control circuits for controlling neutron generator tubes in that improved regulation of the replenisher current is accomplished and a relative simple circuit using few parts is required for this. This replenisher current regulating circuit of the present invention also has a smaller power consumption than those known in the prior art and is capable of operating at temperatures of up to 200.degree. C. An additional feature of the control circuitry of the present invention is the remote turn on or off capability of the neutron generator tube. A further feature of the present invention is the capability of dynamically varying neutron pulse widths and repetition rates or producing dynamically varying neutron output waveforms while maintaining control over the average neutron output of the generator tube.
	abstract	An end support system is applied to each end of a container for shipping nuclear fuel assemblies in an inner box within the container. The end support system includes a rectilinear metal end frame, a crosspiece with a reinforcing channel along an inside surface of the crosspiece and four arms projecting in a perpendicular direction from the metal end frame for straddling the sides of the container end. By screwing the metal end support system into the wooden framing elements of the container, the integrity of the container ends is maintained during hypothetical accident conditions specified in licensing regulations.
	abstract	A control system, and a control method, of a nuclear power plant capable of easily executing automatic output regulation by an automatic power regulator system even when a lower-accuracy maximum linear heat generation rate and a lower-accuracy minimum critical power ratio are determined in a short cycle by using brief calculation. While a core monitoring system does not calculate a maximum linear heat generation rate and a minimum critical power ratio, they are determined by utilizing the maximum linear heat generation rate and the minimum critical power ratio calculated by the core monitoring system as well as plant data, and the lower-accuracy maximum linear heat generation rate and the lower-accuracy minimum critical power ratio so determined are compared with predetermined thermal limit values. The automatic power regulator system holds a control signal when at least one of the lower-accuracy maximum linear heat generation rate and the lower-accuracy minimum critical power ratio exceeds the predetermined thermal limit values.
	claims	1. A device for loading a fuel assembly (5) into a loading location (3) in a core (1) of a nuclear reactor comprising fuel assemblies (5) of general right prismatic shape located within an enclosing wall (2) in adjacent prismatic locations (3) having vertical axes whose transverse cross-sections in a horizontal plane constitute a regular arrangement, a loading location (3) for the fuel assembly comprising at least one vertical lateral surface according to which a lateral surface of a fuel assembly (5) adjacent to the fuel assembly being loaded is positioned, characterized in that it comprises a fuel assembly realignment tool (4) in a form of a dummy assembly having a generally right prismatic shape of a location (3) for a fuel assembly (5) in the core and laterally bounded by smooth walls, andat least one tool for holding fuel assemblies (11) comprising a supporting plate (12) and positioning pins (14) designed to engage simultaneously the positioning openings of upper end members (10) of at least two different fuel assemblies in the core, and to keep said at least two fuel assemblies in relative positions,as well as at least one handling means (15) for gripping and handling the realignment tool for the fuel assemblies (4) and the at least one tool for holding the fuel assemblies (11) through suspension and holding means (8, 13). 2. A device according to claim 1, characterized in that the suspension and holding means (8, 13) of the tool for realigning fuel assemblies (4) and the tool for holding fuel assemblies (11) is similar to a suspension and holding part of an upper end member (10) of a fuel assembly (5) for the core and that the handling device of the tool for realigning fuel assemblies (4) and the tool holding fuel assemblies (11) is a gripper (15) of a nuclear reactor loading machine. 3. A device according to claim 1, characterized in that the tool for realigning fuel assemblies (4) comprises a central body (4a, 4′a), an upper end member (8, 8′) and a lower end member (7, 7′) having a common longitudinal axis (6, 6′) and a transverse cross-section in a plane perpendicular to the axis (6, 6′) having the shape of a transverse cross-section of a location (3) for a fuel assembly in the core of the nuclear reactor. 4. A device according to claim 3, characterized in that the central body (4a) and the lower end member (7) of the tool for the realignment of fuel assemblies (4) of right prismatic shape has a transverse cross-section having dimensions which are smaller than the dimensions of the transverse cross-section of a location for fuel assemblies in the core, an upper end member (8) whose transverse cross-section has the dimensions of the transverse cross-section of a location for a fuel assembly (5) in the core (1) of the reactor and an intermediate part (9) between the central body and the upper end member (8) bounded by lateral walls which are inclined in relation to the axis (6) of the tool for the realignment of fuel assemblies which has a transverse cross-section of generally increasing dimensions between the central body (4a) and the upper end member (8). 5. A device according to claim 3, characterized in that the tool for the realignment of fuel assemblies (4′) comprises a central body (4′a) of right prismatic shape whose transverse cross-section has the dimensions of the transverse cross-section of a location (3) for a fuel assembly (5) in the core (1) of the reactor and a lower end member (7) having lateral walls inclined in relation to the axis (6′) of the tool for the realignment of fuel assemblies in such a way that the transverse cross-section of the lower end member (7′) has decreasing dimensions between the central body (4′a) and its lower engaging extremity in a location (3) in the core (1) of the nuclear reactor. 6. A device according to claim 3, characterized in that the lower end member (7, 7′) of the realignment tool (4) comprises lateral openings (7a) for a passage of positioning pins for a location (3) in the core (1) of the reactor and two posts (26) engaging in the water holes of the location (3) when the realignment tool (4) is positioned on supporting plate for the core (1) of the reactor. 7. A device according to claim 3, characterized in that the lower end member (7) of the realignment tool (4) has a cross-section such that it can be engaged between the positioning pins of a location for a fuel assembly in the core (1) of the reactor and two posts (26′) engaging in the water holes of location (3) when the realignment tool (4) is positioned on the supporting plate for the core (1) of the reactor. 8. A device according to claim 4, characterized in that the walls inclined with respect to the axis (6, 6′) of the intermediate part (9) or the lower end member (7′) of the tool for the realignment of fuel assemblies (4, 4′) have successive portions (9a, 9b) in the direction of the axis (6, 6′) which are inclined with respect to the axis (6, 6′) and substantially parallel to the axis (6, 6′). 9. A device according to claim 2, characterized in that the tool holding the fuel assemblies (11) comprises a supporting plate (12), a first set of positioning fingers (14a) and a second set of positioning fingers (14b) which are parallel to each other and are fixed in arrangements perpendicular to the supporting plate (12), the positioning fingers (14b) of the second set having a length in the direction perpendicular to the supporting plate (12) which is shorter than the length of the fingers (14a) of the first set. 10. A device according to claim 9, characterized in that the positioning fingers (14a) of the first set or long fingers comprise a shank having a first longitudinal axis (14′a) and an extremity tip (22) in a prolongation of the shank having a longitudinal axis (22a) which is offset with respect to the axis (14′a) of the shank in a direction perpendicular to the axis of the shank (14′a). 11. A device according to claim 10, characterized in that the long fingers (14a) of the tool holding fuel assemblies (11) are attached to the supporting plate (12) by mechanical fixing means through which the orientation of a finger (14a) about its longitudinal axis (14′a) and thus the direction of offset between the axis (14′a) of the shank of the finger and (22a) of the extremity tip of the long finger (14a) can be adjusted. 12. A device according to claim 9 for loading fuel assemblies of square transverse cross-section in right prismatic locations of square cross-section in the core (1) of a nuclear reactor, characterized in that the fuel assembly holding tool (11) comprises a supporting plate (12) in the shape of a square whose side is substantially equal to twice the side of the transverse cross-section of one location (3) for a fuel assembly (5) in the core (1) of the nuclear reactor, four long fingers (14a) in the positions of positioning holes for the four adjacent fuel assemblies (5) in the core (1) of the nuclear reactor and four short fingers (14b) of a length shorter than the length of the long fingers in a direction perpendicular to the supporting plate (12) in positions corresponding to the positions in transverse cross-section of four positioning holes for the four adjacent fuel assemblies located on a diagonal on each of the upper end members of the fuel assemblies (5) in relation to the positioning holes in the positions of the long fingers (14a). 13. A device according to claim 9, in the case of a core (1) of a nuclear reactor comprising fuel assemblies (5) having a square transverse cross-section positioned in locations (3) of the core of right prismatic shape having square transverse cross-sections arranged in a square grid arrangement, characterized in that the tool holding fuel assemblies (11′) comprises a supporting plate (12′) in the shape of a square having dimensions corresponding to the dimensions of the transverse cross-sections of the three locations for adjacent fuel assemblies (5) in the core (1), three long pins and three short pins designed to engage respectively the positioning openings of three adjacent fuel assemblies arranged in a square in the core (1) of the nuclear reactor, the long pins being inserted in to the first openings of each of the fuel assemblies and the three short pins being inserted respectively in three second positioning openings for the three fuel assemblies arranged in a square located diagonally with respect to the first openings receiving the long pins. 14. A device according to claim 9, characterized in that the fuel assembly holding tool (11, 11′) comprises a suspension and holding device (13) similar to a suspension and holding device of an upper end member (10) of a fuel assembly (5) integral with the supporting plate (12) on one side of the supporting plate (12) opposite a side of the supporting plate (12) on which the positioning pins (14a, 14b) are fixed projectingly.
	abstract	A cleaner includes an external casing forming a suction hood, an upper suction mouth, the casing also being provided with a drive arranged on each side and equipped with independent motors and corresponding transmission mechanisms on each side, and cleaning rollers; sets of internal cleaning rollers disposed close to the center of the hollow interior of the casing and having a width approximately equal to the distance between the side elements of the casing; sets of external cleaning rollers located close to the front and rear edges of the casing of the cleaning device and having a total width slightly greater than the width of the casing; a resilient joint at the support for the external rollers; a pair of adhesion turbines; auxiliary drive wheels on the internal cleaning rollers; a flotation body connected to the casing and having a fixed and/or variable volume; and laterally mobile turbines.
	abstract	An apparatus for use with an electron beam for imaging a sample. The apparatus has a down-conversion detector configured to detect an electron microscopy signal generated by the electron beam incident on the sample, a direct bombardment detector adjacent to the down-conversion detector and configured to detect the electron microscopy signal, and a mechanism selectively exposing the down-conversion detector and the direct bombardment detector to the electron microscopy signal. A method using the apparatus is also provided.
	abstract	A three-dimensional magneto-optical trap (3D GMOT) configured to trap a cold-atom cloud is disclosed. The 3D GMOT includes a single input light beam having its direction along a first axis, an area along a second and third axis that are both normal to the first axis, and a substantially flat input light beam intensity profile extending across its area. The 3D GMOT may also include a circular, diffraction-grating surface positioned normal to the first axis and having closely adjacent grooves arranged concentrically around a gap formed in its center. The circular, diffraction-grating surface is configured to diffract first-order light beams that intersect within an intersection region that lies directly above the gap and suppresses reflections and diffractions of all other orders. The 3D GMOT may further include a quadrupole magnetic field with its magnitude being zero within the intersection region.
061817732	abstract	A radiation anti-scatter device comprising a grid and a grid driver connected to the grid for unidirectionaly moving the grid with a variable grid velocity along a path between a starting and an end position, and a method of providing such grid motion. The variable grid velocity may have a velocity profile V.sub.1 =k.sub.1 t for a first period and then V.sub.2 =k.sub.2 t.sup.-m for a second period, where V.sub.1 and V.sub.2 are velocity, k.sub.1 and k.sub.2 are constants, t is time, and m is an exponent having a value greater than 0. The anti-scatter device may be a component of a direct radiographic diagnostic imaging system which includes an image-producing element having an array of radiation detectors aligned in rows, and where the anti-scatter device is a grid having vanes oriented at an angle to the detector rows. Radiation emission may be synchronized with the grid motion to optimize a radiograph for a particular grid, radiation source, or examination procedure. The apparatus implements a method for reducing Moire patterns in radiographic detectors having an array of sensors by unidirectionaly moving the grid in a single stroke during the radiation exposure with an asymptotically decreasing speed profile such that grid motion is maintained for a plurality of different radiation exposure times.
	abstract	Disclosed are embodiments of an ion beam sample preparation apparatus and methods for using the embodiments. The apparatus comprises an ion beam irradiating means in a vacuum chamber that may direct ions toward a sample, a shield blocking a portion of the ions directed toward the sample, and a shield retention stage with shield retention means that replaceably and removably holds the shield in a position. The shield has datum features which abut complementary datum features on the shield retention stage when the shield is held in the shield retention stage. The shield has features which enable the durable adhering of the sample to the shield for processing the sample with the ion beam. The complementary datum features on both shield and shield retention stage enable accurate and repeatable positioning of the sample in the apparatus for sample processing and reprocessing. Additionally, apparatus kits are disclosed that enable the use of the same shields in the observation of prepared samples.
048427748	claims	1. An improved waste disposal site for the above-ground disposal of radioactive wastes which is formed from a plurality of individual, waste-containing tumuli, wherein each tumuli includes a central raised portion bordered by a sloping side portion, comprising at least two ground-level tumuli having mutually adjoining side portions, and an above-ground tumulus disposed over the adjoining side portions, wherein each tumulus includes a deformable floor and roof in order to render each tumulus flexibly conformable with the surrounding terrain in the event of a seismic disturbance, and the roofs of the ground-level tumuli include both an intrusion barrier formed from a plurality of flexibly interlocking structures, and a deformable layer of water-shedding material that forms at least part of the floor of the above-ground tumulus, and wherein each tumulus contains at least one shield wall means for protecting workers from potentially harmful radiation when said tumulus is constructed. 2. An improved waste disposal site as defined in claim 1, wherein each of said interlocking structures includes a plurality of legs, and a connector member for interconnecting said legs. 3. An improved waste disposal site as defined in claim 1, wherein each tumulus contains radioactive wastes of greater and lesser radioactivity which are positioned on different sides of said shield wall means. 4. An improved waste disposal site as defined in claim 1, wherein the slope of each of the side portions of the tumuli is at least one to four. 5. An improved waste disposal site as defined in claim 1, wherein said shield wall means is formed from a flexibly conformable array of wall components. 6. An improved waste disposal site as defined in claim 5, wherein each of said wall components is a container that holds nuclear wastes, wherein the radiation intensity at the surface of each container is no greater than about 100 millirems/hour. 7. An improved waste disposal site for the above-ground disposal of radioactive wastes which is formed from at least three individual, waste-containing tumuli, wherein each tumulus has a deformable floor and roof for rendering each tumulus flexibly conformable with the surrounding terrain in the event of a seismic disturbance, each of which includes a water-shedding, deformable layer of material, said roof also including an intrusion barrier formed from a plurality of flexibly interlocking structures, and the water-shedding, deformable layer of material of the roof of one tumulus forms at least in part the water-shedding, deformable layer of material of the floor of another tumulus, and wherein each tumulus contains at least one shield wall means for protecting workers from potentially harmful radiation when said tumulus is constructed. 8. An improved waste disposal site as defined in claim 7, wherein the roof of each tumulus includes a central raised portion bordered by a sloping side portion. 9. An improved waste disposal site as defined in claim 8, wherein the slope of each said side portion is about one to four. 10. An improved waste disposal site as defined in claim 7, wherein each of said structures includes a plurality of legs which interfit with the legs of adjacent structures. 11. An improved waste disposal site as defined in claim 7, wherein each tumulus contains radioactive wastes of greater and lesser radioactivity which are positioned on different sides of said shield wall means. 12. An improved waste disposal site as defined in claim 7, wherein said water-shedding deformable layer of both said roof and said floor is formed from compacted clay. 13. An improved waste disposal site as defined in claim 7, wherein each floor further includes a layer of water-permeable material that contains natural zeolites to retard the passage of radioactive chemicals through said layer. 14. An improved waste disposal site as defined in claim 13, further including a plurality of drainage conduits disposed at the bottom of said layer of water-permeable material for collecting and directing water to a drainage gallery means. 15. An improved waste disposal site for the above-ground disposal of radioactive wastes comprising a plurality of individual, waste-containing tumuli, each tumulus including a deformable floor and deformable roof for rendering each tumulus flexibly conformable with the surrounding terrain in the event of a seismic disturbance, wherein said floor is formed in part by a deformable layer of water-shedding material, and said roof is likewise formed in part by a deformable layer of water-shedding material, and said roof further includes an intrusion barrier formed from a plurality of flexibly interlocking structures and has a central raised portion, wherein said site includes at least two ground-level tumuli having mutually adjoining side portions, and an above-ground tumulus disposed over the adjoining side portions whose floor is formed in part by the water-shedding deformable layer of material in the roofs of the ground level tumuli, and wherein each tumulus contains at least one shield wall means for protecting workers from potentially harmful radiation when said tumulus is constructed. 16. An improved waste disposal site as defined in claim 15, wherein each floor further includes a layer of water-permeable material disposed on top of its water-shedding deformable layer of material, wherein said water permeable layer contains natural zeolites to retard the passage of radioactive chemicals through said layer. 17. An improved waste disposal site as defined in claim 16, further including a plurality of drainage conduits disposed at the bottom of said layer of water-permeable material for collecting and directing water to a drainage gallery means. 18. An improved waste disposal site as defined in claim 17, further including a radiation detecting means for detecting the level of radiation of said water in said gallery means. 19. An improved waste disposal site as defined in claim 18, wherein said gallery means includes a valve means for diverting the water therein to a treatment plant in the event the radioactivity of the water collected in the gallery means exceeds a selected value. 20. A pyramidal array of tumuli for the above-ground disposal of radioactive wastes, comprising at least two ground-level tumuli and one above-ground tumulus, each having a deformable floor and a roof for rendering each tumulus flexibly conformable with the surrounding terrain in the event of a seismic disturbance, wherein each said floor and roof includes a deformable layer of water-shedding material, and said roof of each includes a ceiling formed from another deformable layer of water-shedding material, and an intrusion barrier disposed between said deformable layers of water-shedding material in said roof and ceiling that is formed from a plurality of flexibly interlocking structures, as well as a central raised portion bordered by a sloping side portion which is supported by an array of nuclear waste packages, and wherein each tumulus contains a shield wall means for protecting workers from potentially harmful radiation during the construction of the array. 21. A pyramidal array of tumuli as defined in claim 20, wherein the floor of each tumulus includes a layer of water-permeable material for supporting said array of nuclear waste packages and draining any water that should collect around said packages. 22. A pyramidal array of tumuli as defined in claim 21, wherein each tumulus includes a plurality of drainage conduits disposed at the bottom of said layer of water-permeable material for collecting and directing water to a drainage gallery means. 23. A pyramidal array of tumuli as defined in claim 22, wherein each tumulus includes a radiation detecting means for detecting the level of radiation of said water in said gallery means, as well as a valve means for diverting the water in said gallery means to a treatment plant in the event that the radioactivity of the water collected in the gallery means exceeds a selected value. 24. A method of constructing a site for the above-ground disposal of radioactive wastes, comprising the steps of a. constructing a deformable floor for a first tumulus that includes a deformable layer of water-shedding material; b. constructing a shield wall means on said floor capable of blocking radiation from said wastes; c. depositing an array of packages containing radioactive wastes over said floor and behind said shield wall; d. constructing a deformable roof over the array of waste packages which includes a deformable layer of water-shedding material, and an intrusion barrier beneath the deformable layer of water-shedding material, said barrier being constructed by overlaying the array of waste packages with a plurality of flexibly interlocking structures; e. constructing a second tumulus in accordance with steps a-d which is close enough to the first tumulus so that the deformable layers in the roofs of the first and second tumuli adjoin one another; and f. constructing a third tumulus over the roofs of the first and second tumuli in accordance with steps a-d wherein the mutually adjoining, deformable layers of water-shedding material in said roofs provide the deformable layer of water-shedding material of the floor of the third tumulus. 25. A method of constructing a site for the above-ground disposal of [toxic] radioactive wastes as defined in claim 32, wherein said waste packages contain radioactive wastes, and the surface radiation intensity of some of said packages is higher than others. 26. A method of constructing a site for the above-ground disposal of [toxic] radioactive wastes as defined in claim 25, wherein [the disposal operators load] the waste packages of higher surface radiation intensity are loaded behind the shield wall means when [depositing] said array of packages are deposited over said tumulus floor in order to minimize [their exposure to] the amount of radiation radiated toward persons in the area.
	abstract	An X-ray generator, an X-ray inspector and an X-ray generation method capable of automatically focusing an energy beam, such as an electron beam for generating an X-ray, on a target are provided. The generation, inspector and the method have been developed by turning an attention on the fact that convergence conditions of an electron beam has a close relationship with a temperature on a surface of an X-ray tube target. The method comprises the steps of measuring the temperature changes at real time by a temperature sensor 14 and automatically controlling a current value of a focusing coil 6.
047537728	claims	1. An energy dissipating tension loaded support member comprising: a pair of spaced apart end members to which a tension load is applied; and a plurality of straps of successively longer lengths of stiff metallic material each connected at each end to an end member such that at least all of said straps but the shortest one are bowed with no load applied to the end members, the lengths of said straps being selected such that beginning with the shortest strap, the successively longer straps sequentially reach their yield points and plastically deform to dissipate energy as the tension load on said end members increases. a pair of spaced apart end members to which a tension load is applied; three metal straps, the second metal strap being longer than the first and the third being longer than the second; and means connecting one end of each of the three metal straps to one end member and connecting the other ends of said straps to the other end member, such that at least said second and third straps are bowed when no tension load is applied to the end members. a nuclear reactor, a steam generator, piping connecting the nuclear reactor with the steam generator for circulation of reactor coolant therebetween, and seismically qualified supports for said piping at selected points along the length thereof each comprising a plurality of energy dissipating tension loaded support members angularly distributed about and extending radially outward from said piping, said tension loaded support members each comprising a pair of spaced apart end members, one connected to said piping and the other to a fixed support, and a plurality of straps of successively longer lengths of stiff metallic material each connected at each end to an end member such that at least all of said straps but the shortest one are bowed with no load applied to the end members, the lengths of said straps being selected such that beginning with the shortest strap, the successively longer straps sequentially reach their yield points and plastically deform to dissipate energy as the tension load on said end members increases. 2. The support member of claim 1 wherein said shortest strap is also bowed when no load is applied to the end members. 3. The support member of claim 1 wherein said straps are selected such that the successively longer straps require successively higher loads to plastically deform. 4. The support member of claim 1 wherein at least said shortest strap yields to the point of rupture under expected tension loads. 5. The support member of claim 4 wherein at least the longest strap remains elastic with the expected tension loads. 6. The support member of claim 3 wherein said straps are all made of the same metallic material with successively longer straps having successively greater cross-sectional areas. 7. The support member of claim 6 wherein said straps are all of the same thickness with successively longer straps being successively wider. 8. The support member of claim 7 comprising three successively longer and wider straps. 9. The support member of claim 8 wherein the shortest strap is also bowed when no load is applied to the end members. 10. A multi-strap shock absorber comprising: 11. The shock absorber of claim 10 wherein said three straps are made of the same metal and are of the same thickness, said second strap being wider than the first strap and the third strap being wider than the second. 12. The shock absorber of claim 11 wherein said first strap is also bowed when no tension load is applied to said end members. 13. A nuclear steam supply system comprising: 14. The nuclear steam supply system of claim 13 wherein said seismically qualified supports for said piping comprise three tension loaded support members angularly distributed 120.degree. apart around said piping and extending radially outwardly therefrom. 15. The nuclear steam supply system of claim 14 wherein at least said longest strap of each tension loaded support member remains elastic even with the largest seismic load expected.
062460528	abstract	A flexure carriage assembly has a carriage formed of a substantially rigid material. The carriage has four elongate columns arranged spaced apart and parallel to one another. Each of the elongate columns has first and second ends. The flexure carriage has four first cross members disposed between adjacent pairs of elongate columns and arranged to interconnect the first ends. The flexure carriage also includes four second cross members arranged between adjacent pairs of elongate columns and arranged to interconnect the bottom ends. The elongate columns and first and second cross members define a three-dimensional rectangular structure. The flexure carriage also has disposed centrally between the four elongate columns a translating section spaced equidistant between the first and second ends of the columns. A plurality of flexures are disposed between the translating element and elongate columns and between the elongate columns and first and second cross members in order to permit precise movement of the translating section in a plane according to applied forces against edges of the translating section. A pair of piezoelectric assemblies are connected to the translating section. One applies force to the translating section in a first linear path and the other applies force to the translating section in a second linear path perpendicular path.
	description	The present application is a divisional of co-pending U.S. patent application Ser. No. 11/792,622, filed Jun. 7, 2007, which is a national phase of International Application No. PCT/FR2005/050841 entitled “Device For Cleaning And/Or Securing A Containment Chamber Defined In A Device Used For The Transport And/Or Storage of Radioactive Materials”, which was filed on Oct. 12, 2005, which was not published in English, and which claims priority of the French Patent Application No. 04 52884 filed Dec. 7, 2004. This invention generally relates to the field of cleaning and/or securing a safe containment defined in a device for transporting and/or storing radioactive materials, such as, for example, nuclear fuel assemblies. In such safe containments holding nuclear fuel assemblies, the high temperature promotes the conversion of oxides stable under ambient conditions into water, and the radiation emitted by the fuel assemblies cause some of this water to be converted by radiolysis into hydrogen and oxygen. Thus, because the presence of hydrogen in the containment clearly jeopardizes the safety of the transportation and/or storage device assembly due to significant risks of flammability, explosiveness and pressure build-up that it creates, it is routinely attempted to remove this hydrogen. To do this, the safe confinements can be equipped with a catalyst for recombining oxygen and hydrogen into water (or catalytic hydrogen recombiner), in contact with which the hydrogen combines with the oxygen present in the safe containment to form water according to the catalytic oxidation mechanism of hydrogen. Naturally, the presence of water inside this containment also creates significant risks of corrosion, so that it may be necessary to clean the containment, in particular to remove the water, for example using a drying agent. Of course, it is noted that damaging elements other than those cited above may be present in the containment, which may lead to the use of additional active means suitable for cleaning and/or securing this same safe containment. The prior art includes a first document EP-A-0 660 335, in which active means for cleaning and/or securing the safe containment are contained in a sealed cavity insulated by a bursting membrane, of which the rupture is caused by the development of overpressure inside the safe containment. An advantage related to this solution lies in the fact that the overpressure causing the burst of the membrane can occur only after the containment has been drained of borated water, meaning that the active means are therefore never in contact with the borated water capable of rendering them inoperative, temporarily or permanently. Nevertheless, this solution has a plurality of major disadvantages. Indeed, it is first indicated that the active means enabling the safe containment to be cleaned and/or secured are only in contact with the environment of this containment by means of an opening corresponding to the burst membrane. Thus, significant convection currents cannot be produced around the active means, so that the overall efficacy of the latter remains relatively low. In addition, when nuclear fuel assemblies are packaged, it is of course necessary to perform an additional pressurisation operation so as to generate a burst of the membrane, which is susceptible at this time of being broken and scattered in the safe containment. The prior art also includes a second document EP-A-0 895 250, in which the safe containment is equipped with a residual moisture absorber with a molecular sieve, arranged vertically below a closable opening formed in a cover of the transportation and/or storage device. In addition, the size of this opening is such that it enables the residual moisture absorbent to be introduced into the safe containment, and removed therefrom. Although this other solution also enables the risks associated with the incompatibility between the active means and the borated water of the pool inside the safe containment to be avoided when filling fuel assemblies, it is not entirely satisfactory, in particular doe to the fact that it is implemented after the final packaging of the cavity and necessitates an additional operation. The invention therefore aims first to propose a device for cleaning and/or securing a safe containment defined in a device for transporting and/or storing radioactive materials such as nuclear fuel assemblies, which device at least partially overcomes the disadvantages mentioned above with regard to the devices of the prior art. More specifically, the objective of the invention is to propose a device for cleaning and/or securing a safe containment which has satisfactory effectiveness, a simple design, and is capable of ensuring protection of the active means from the borated water in the safe containment during the loading of fuel assemblies. In addition, the invention also aims to provide a device for transporting and/or storing radioactive materials such as nuclear fuel assemblies including at least one device for cleaning and/or securing, as well as a method for underwater packaging of radioactive materials also using at least one such device for cleaning and/or securing the safe containment. To do this, the invention first relates to a device for cleaning and/or securing a safe containment defined in a device for transporting and/or storing radioactive materials such as nuclear fuel assemblies, which device includes active means capable of cleaning and/or securing the safe containment. According to the invention, this device also includes: means forming a casing with an opening and defining a cavity; and means for closing the opening, capable of assuming an open position as well as a closed position in which they close this opening so as to seal the cavity in which the active means are located, which device is designed so that the closure means are capable of being held in the closed position by a pressure difference having a value greater than or equal to a predetermined value, between the inside and the outside of the cavity. In other words, it should be understood that the device according to the invention is such that the active means can be enclosed in a sealed cavity by maintaining the closure means in the closed position, solely by applying a simple pressure difference between the inside and the outside of said cavity, preferably obtained by creating a vacuum therein resulting in a pressure inside the cavity that is lower than that outside said cavity. Thus, in this closed state, the device can then be located in the safe containment even when the borated water is still present therein, without the risk of the active cleaning and/or securing means being rendered inoperative by said borated water. In addition, in order for these active cleaning and/or securing means to satisfy their function in the safe containment which has previously been drained, and therefore emptied of its borated water, it is sufficient to eliminate the pressure difference having a value greater than or equal to the predetermined value, in order to cause the closure means of said cleaning and/or securing device to automatically switch from the closed position to the open position. Indeed, it is naturally specified that when the closure means assume the open position, the open cavity then enables the active means to communicate directly with the atmosphere of the safe containment. By way of indication, it is noted that the predetermined pressure difference value mentioned above is in particular based on the design of the device, and that it can, for example, correspond to a minimum value not only maintaining the closure of the opening of the cavity by the closure means, but also a satisfactory seal between the means forming a casing with an opening, and these same closure means. In this regard, it is indicated that the predetermined value does not necessarily have to be set so as to cause an automatic movement of these closure means from the open position to the closed position, which manipulation can actually be performed manually by an operator, before the pressure difference between the inside and the outside of the cavity is applied, so as to enable this closed position to be maintained, and thus provide a containment for the active means. In addition, it has been noted that the rupture from the pressure difference with a value greater than or equal to the predetermined value caused the closure means to automatically switch from the closed position to the open position. Thus, this rupture refers not necessarily to a total rupture from the pressure difference between the inside and the outside of the cavity, but to a decrease in this pressure difference to a value leading to the desired result, namely the automatic release of the opening of the cavity. Naturally, if, as was just described, the design of the device is such that the release of the opening can be achieved before the pressure difference between the inside and the outside of the cavity reaches zero, it is clear that this automatic release also occurs when the pressure difference is reversed, namely when the pressure outside the cavity is lower than the pressure inside the cavity. The cleaning and/or securing device according to the invention was designed so that the open position of the closure means is achieved automatically during the operation of drying the safe containment following the draining operation, which drying operation is indeed performed by creating a depression in the containment, in which the pressure is reduced and approaches that of the cavity still under pressure. Once the closure means are in open position, the active means are arranged in the core of the safe containment, which enables their efficacy to be enhanced with respect to that found in prior art document EP-A-0 660 335. In addition, as is clear from the above, its design and operation remain relatively simple, and the active cleaning and/or securing means are entirely protected from the borated water in which the device can be immersed in a closed configuration. Finally, it is specified that another advantage lies in the fact that the design and operation of this device enable it to be easily refit. The device can preferably be designed so that the closure means switch from the closed position to the open position automatically by means of gravity, after a rupture from the pressure difference having a value greater than or equal to the predetermined value. Moreover, in this specific case, after the aforementioned automatic switch, the closure means are preferably maintained in the open position also by gravity. According to another alternative, the device can be designed so that the closure means switch from the closed position to the open position automatically by way of elastic means inserted between the closure means and the means forming a casing, after a rupture from the pressure difference having a value greater than or equal to the predetermined value. However, in this specific case, it is clear that the automatic switch is triggered when the internal pressure is still lower than the external pressure. In addition, after the aforementioned automatic switch, the closure means are preferably maintained in the open position also by way of elastic means. Of course, it is possible to envisage providing a device combining gravity and the elastic means so as to enable the switch from the closed position to the open position, without going beyond the scope of the invention. The device preferably includes means enabling a depression to be created in the cavity. Also preferably, the active means are mounted on the closure means so that when the latter assume the open position in which they are located at a distance from the opening, the active means are located at least partially outside the cavity. Naturally, this particular arrangement enables the active means to have an increased overall efficacy in the safe containment. The closure means preferably comprise holding means enabling said closure means to be maintained in the open position. In addition, the holding means comprise a pin capable of sliding inside a hollow cylinder secured to the means forming a casing with an opening, which pin has a shoulder located inside the hollow cylinder and capable of coming into contact with an abutment provided at one end of said hollow cylinder. In such a configuration, the pin can be attached to a closure plug bearing the active means. Thus, during the switch from the closed position to the open position, the closure plug and the active means are moved simultaneously. Of course, this applies only when the means forming a casing remain stationary with respect to the containment and the closure means are moved with respect to these same means forming the casing during the automatic switch between the two positions, and not in the reverse case, which can also naturally be envisaged. Also preferably, the active means are arranged in one or more cartridges, and can include a catalyst for recombining oxygen and hydrogen into water, as well as a drying agent. By way of illustration, these two elements can be combined and take the form of palladium deposited on alumina, enabling the catalysis as well as the required drying to be achieved simultaneously. The invention also relates to a device for transporting and/or storing radioactive materials such as nuclear fuel assemblies, including at least one cleaning and/or securing device such as the one that has just been described, which device is located in the safe containment and the closure means assume the open position. Finally, the invention also relates to a method for underwater packaging of radioactive materials such as nuclear fuel assemblies, including the following steps: creating a vacuum in the cavity of a cleaning and/or securing device as described above, so that the closure means are maintained in the closed position; mounting the cleaning and/or securing device inside the safe containment of a device for transporting and/or storing radioactive materials; placing the transportation and/or storage device in a pool after filling the safe containment with water; loading the radioactive materials into the safe containment; closing the transportation and/or storage device using at least one cover; extracting the transportation and/or storage device from the pool; draining the water located inside the safe containment; and drying said safe containment by creating a depression in the latter, which depression is created so as to cause the closure means of the cleaning and/or securing device to automatically switch from the closed position to the open position. Thus, as described above, it should be understood that there is a correlation between the value of the depression applied to obtain the vacuum and the closure of the cavity, and the value of the depression applied in the safe containment to cause the latter to dry out, which correlation is determined so that the pressure difference having a value greater than or equal to the predetermined value between the inside and the outside of the cavity, enabling the closure means to be maintained in the closed position, is sufficiently attenuated, or even reduced to zero or else reversed during the drying, so as to allow the closure means of the cleaning and/or securing device to automatically switch from the closed position to the open position. By way of example, the value of the two depressions indicated above can be identical. Other advantages and characteristics of the invention will appear in the non-limiting detailed description below. First, FIGS. 1 to 3 show a device 1 for cleaning and/or securing a safe containment defined in a device (not shown in FIGS. 1 to 3) for transporting and/or storing radioactive materials such as nuclear fuel assemblies, according to a preferred embodiment of this invention. The device 1 comprises means forming a casing 2 with an opening, which preferably have the shape of an annular body 4 closed at one of its two ends by a cover 6, and defining an opening 8 at the other of its ends. In addition, the annular body 4 defines a cavity 9 open only at the level of the opening 8. In the following description of FIGS. 1 to 3, by convention, the cover 6 is considered to be at the level of a high end of the annular body 4, and the opening 8 is considered to be at the level of a low end of this same annular body. The device 1 also includes means for closing 10 the opening, which means 10 primarily consist of a closure plug 12 oriented so as to be substantially perpendicular to a main longitudinal axis 14 of the means forming the casing 2. The plug 12 can be positioned so as to close the opening 8 as shown in FIG. 1, therefore showing the closure means 10 in a closed position, and can also be located at a distance from this same opening 8 as shown in FIG. 3 showing the closure means 10 in an open position. Thus, the closure means 10 comprise holding means 16 enabling the closure means 10 to be maintained in the open position, which holding means 16 thus enable a mechanical link to be maintained between the closure plug 12 and the means forming the casing 2, when said plug 12 is located at a distance from the opening 8. More specifically, the holding means 16 comprise a pin or solid cylinder 20 parallel to the main longitudinal axis 14 of the body 4, which pin 20 is capable of sliding into a hollow cylinder 22 also oriented according to the main longitudinal axis 14, and is secured to the cover 6 of the means forming the casing 2. Preferably, as is clearly visible in FIGS. 1 to 3, the pin 20 and the hollow cylinder are centred on the main longitudinal axis 14 of the body 4. In this regard, the pin 20 bears the plug 12 at one of its ends as well as a shoulder 24 at the other of its ends, which shoulder 24 is located inside the hollow cylinder 22 and is capable of coming into contact with an abutment 26 provided at a low end of said cylinder 22. In this way, as can be seen in FIG. 3, when the closure means 10 assume the open position, the closure plug 12 is maintained at a distance from the opening 8 and perpendicular to the main longitudinal axis 14 by the contact between the shoulder 24 and the abutment 26. In addition, as the abutment 26 of the hollow cylinder 22 is located near this opening 8, practically the entire pin projects downward when the closure means 10 assume the open position. Also in reference to FIGS. 1 to 3, the device 1 has active means capable of cleaning and/or securing the safe containment, which active means are, for example, arranged in one or more cartridges 28, which are secured at one of their ends to the closure plug 12. By way of example, and as is best shown in FIG. 2, the cartridges 28 are preferably arranged parallel to the main longitudinal axis 14 around the pin 20 of the holding means 16, so as to be capable of being inserted into the cavity 9 when switching the closure means 10 from the open position to the closed position. Thus, it can actually be seen that in the closure position of FIG. 1, the cartridges 28 are located between the annular body 4 and the hollow cylinder 22, in which annular cavity 9 the opening 8 is hermetically closed by the closure plug 12 secured to the low ends of these cartridges 28. However, in the open position shown in FIG. 3, the cartridges 28 project downward with respect to the annular body 4, and thus each have at least a portion located outside the cavity 9, intended to be arranged at the core of the safe containment to be cleaned and/or secured. The active cleaning and/or securing means present in the cartridges 28 preferably include a catalyst for recombining oxygen and hydrogen into water as well as a drying agent, for the reasons mentioned above in the prior art section. By way of example, the drying agent is selected from silica gel, molecular sieves, dehydrated complexing agents such as, for example, copper sulphate or hygroscopic chemical products such as calcium chloride, magnesium sulphate, or phosphorus pentoxide, possibly on a support material. The recombination catalyst is selected in particular from platinum- or palladium-coated catalysts. Also by way of preferred example, the active means can take the form of palladium deposited on alumina, enabling the catalysis and the required drying to take place simultaneously. Naturally, these active means are determined and retained according to the nature of the elements to be removed inside the safe containment of the transportation and/or storage device, so as to clean and/or secure this same containment. With device 1, when the closure means 10 assume the open position in which the closure plug 12 is held by gravity at a distance from the opening 8, the closure position can then be obtained by exerting a simple manual action so as to translate said plug 12 until it comes into contact with the annular body 4 and closes the opening 8 of the cavity 9. Then, for this closure position to be maintained regardless of the orientation of the device 1 in space, and therefore so that the closure plug 12 does not release the opening 8 under the effect of gravity, a pressure difference having a value greater than or equal to a predetermined value is applied between the inside and the outside of the cavity 9. The pressure difference mentioned above is achieved by creating a vacuum in the cavity 9, by means of an opening 30 provided for this purpose on the cover 6 of the means forming the casing 2, as well as using pumping means (not shown). By way of example, the predetermined pressure difference value can, for example, be set at around 850 mbar, and is in every case maintained so that the contact between the closure plug 12 and the annular body 4 is impermeable, so that the active means arranged inside the cavity 9 are not disturbed by the external environment. Thus, the pressure difference having a value greater than or equal to the predetermined value can be obtained by creating a vacuum in the cavity 9 generating a pressure of around 150 mbar inside the latter. In addition, when the device 1 is properly positioned, it is possible to automatically return to the open position of the closure means 10 by gravity by causing a change in the pressure difference having a value greater than or equal to the predetermined value on each side of the closure plug 12. Indeed, the creation of a depression in the means forming the casing 2, which are stationary with respect to the safe containment, results in a decrease in the pressure difference between the inside and the outside of the cavity 9, and causes the plug 12 as well as the cartridges 28 to automatically fall due to gravity, until they are held in the open position by the contact between the shoulder 24 and the abutment 26. FIG. 4 shows the device 1 in an alternative form of the preferred embodiment described above in reference to FIGS. 1 to 3. Consequently, the elements with the same numeric references correspond to identical or similar elements. Thus, it is indicated that the only difference between these two devices 1 lies in the fact that the one shown in the alternative form comprises elastic means inserted between the means forming a casing 2 and the closure means 10, which elastic means are designed so as to generate the automatic switch from the closed position to the open position, after a rupture from the pressure difference having a value greater than or equal to the predetermined value. The elastic means preferably have the form of a simple compression spring 34 contacting an inner surface of the cover 6 as well as an inner surface of the closure plug 12, which spring 34 is, for example, located around the hollow cylinder 22. With such an arrangement, as the automatic switch to the open position is no longer caused, at least exclusively, by gravity, but by way of elastic means, the device 1 is then advantageously capable of functioning satisfactorily regardless of the orientation of the latter inside the safe containment. To show this, as can be seen in FIG. 4, the device 1 can indeed be in a turned-over position with respect to that of FIGS. 1 to 3, namely with the cover 6 at the level of a low end of the annular body 4, and with the opening 8 at the level of a high end of this same annular body. In this regard, it is noted that the elastic means also enable the closure means 10 to be maintained in the open position in which the cartridges 28 are located outside the cavity, and thus prevent these means 10 from falling due to gravity into the closed position. Naturally, as mentioned above, the presence of elastic means indicates that the automatic switching of the closure means 10 is activated when the pressure difference on each side of the plug 12 is not yet zero, i.e. at a time when the pressure inside the cavity 9 is still lower than the pressure outside the latter. FIGS. 5a to 5k show different steps of a method for underwater packaging of nuclear fuel assemblies, according to a preferred embodiment of this invention using one or more devices 1 such as that shown in FIGS. 1 to 3. Naturally, the packaging method could also be implemented using one or more devices 1 such as that shown in FIG. 4 in the open position, without going beyond the scope of the invention. In addition, for reasons of clarity of the description, the method will hereinafter be considered to require the presence of only a single device 1 to clean and/or secure the safe containment. First, in reference to FIGS. 5a to 5c, the closure means 10 of the device 1 are placed manually from the open position to the closed position, then the cavity 9 is subjected to a vacuum by means of the opening 30, until the pressure difference on each side of the closure plug 12 is greater than or equal to the predetermined value, ensuring that the closure means 10 are locked in closed position, and that the active means are contained inside the cavity 9. The device 1 can then be assembled inside the safe containment 102 of a device 100 for transporting and/or storing nuclear fuel assemblies to be packaged, before said device 100 is immersed in a pool. By way of example, the device 1 can be mounted on a storage bin 104 arranged inside the containment 102, on a container 106 laterally defining this same containment, or a cover 108 closing the latter. In the description below, as is visible in the Figures, the device 1 is considered to be mounted on the bin 104. Regardless of the choice made for mounting the device 1, as the opening after the latter is intended to occur automatically by way of gravity, the orientation of the device 1 inside the containment 102 is preferably that shown in FIGS. 1 to 3, namely with the closure plug 12 downward and the cover 6 upward, and the device 100 in a position in which it rests on its base (not shown). Of course, in such a case in which the main longitudinal axis 14 of the device 1 is intended to be parallel to a main longitudinal axis 103 of the device 100, it is therefore the means forming the casing 2 that are secured to the bin 104. Once the device 1 is assembled, the cover(s) 108 of the device 100 are mounted on the container 106 as shown in FIG. 5d, and the transportation and/or storage device assembly 100 can then be immersed in a pool 109 as shown diagrammatically in FIG. 5e, after the safe containment 102 has been filled with water. In FIG. 5f, it can be seen that the cover(s) 108 are again removed so as to enable the nuclear fuel assemblies 110 to be loaded into the safe containment 102, in the storage bin 104 holding the device 1, as is clear from FIG. 5g. Then, when the loading of the assemblies 110 has been completed, the container 106 is again closed using the cover(s) 108 as shown in FIG. 5h, with the device 1 still located integrally in the pool 109, and the cleaning and/or securing device 1 still in its closed configuration protecting the active means from the borated water present in the pool 109. The device 100, of which the safe containment 102 is now hermetically closed and filled with water, is then moved so as to be extracted from the pool 109, as shown diagrammatically in FIG. 5i. Next, a step of draining the water located inside the safe containment 102 is performed in a manner known to a person skilled in the art. Also in a known manner, a step of drying this same safe containment 102 is then performed, by creating a depression in the latter, for example, by means of an opening (not shown) provided for this purpose in the cover 108. The special feature of the packaging method lies in the fact that the depression created during the drying is performed so as to simultaneously cause the closure means 10 of the device 1 to automatically switch from the closed position to the open position. Indeed, the value of the depression applied inside the cavity 9 to obtain the vacuum thereof and the value of the depression applied in the safe containment 102 to ensure that it dries are set on the basis of one another so that during the drying, the pressure difference between the inside and the outside of this cavity 9 is sufficiently attenuated to cause the aforementioned automatic switch, by means of gravity. By way of example, the two aforementioned depression values are each set at around 150 mbar. Thus, as shown in FIG. 5j, the device 100 has a safe containment 102 equipped with a cleaning and/or securing device 1 in an open configuration, i.e. in a configuration in which the active means are directly in contact with the atmosphere of said containment 102. To conclude, the method can comprise standard steps such as exposing the safe containment 102, or the protective cover installation 112 at the upper and lower ends of the device 100, to inert gas, such as helium. It is noted that the invention also relates to a device 100 for transporting and/or storing radioactive materials, such as that shown in FIG. 5j in which the device 1 includes closure means in the open position. Finally, it is noted by way of indication that, in the case of wet transport of nuclear fuel assemblies, the device 1 is assembled inside the safe containment so as to always be located in the gaseous portion of the latter and not in the liquid portion thereof, naturally so as to preserve the efficacy of the active cleaning and/or securing means. Obviously, this constraint does not apply in the case of dry transport. Various modifications can of course be made by a person skilled in the art to the devices 1, 100 and to the packaging method described above solely by way of non-limiting examples. In this regard, it is noted that the invention applies not only to nuclear fuel assemblies as presented above, but also to any other type of radioactive material.
	claims	1. A sight device, in particular a reflector sight or telescopic sight, which comprises a lighting apparatus for generating or illuminating a target mark,wherein the lighting apparatus comprises a light guide made of photoluminescent, in particular fluorescent material and a radioluminescent light source coupled to the light guide,wherein the light guide is designed to receive both the light produced by the radioluminescent light source as well as ambient light,wherein the light guide is configured to receive the ambient light along at least a section of a longitudinal extension of the light guide and to convert the ambient light into photoluminescent light,wherein the absorption spectrum of the photoluminescent material of the light guide and the emission spectrum of the radioluminescent light source in the visible range can both comprise a spectral bandwidth and a center wavelength,wherein the center wavelength of the emission spectrum of the radioluminescent light source is greater than the center wavelength of the absorption spectrum of the photoluminescent material of the light guide, andwherein the radioluminescent light source is covered by an opaque coating which reflects the light produced by the radioluminescent light source back to the radioluminescent light source, and the opaque coating is applied onto a surface of the radioluminescent light source. 2. The sight device as claimed in claim 1, wherein the center wavelength of the emission spectrum of the radioluminescent light source is at least 30 nm, preferably at least 50 nm, greater than the center wavelength of the absorption spectrum of the photoluminescent material of the light guide. 3. The sight device as claimed in claim 1, wherein the spectral bandwidth of the emission spectrum of the radioluminescent light source and the spectral bandwidth of the absorption spectrum of the photoluminescent material of the light guide each amounts to at most 100 nm, preferably at most 80 nm. 4. The sight device as claimed in claim 1, wherein the spectral bandwidth of the emission spectrum of the radioluminescent light source and the spectral bandwidth of the absorption spectrum of the photoluminescent material of the light guide do not overlap. 5. The sight device as claimed in claim 1, wherein in the visible range at most 30%, preferably at most 20%, of the emission spectrum of the radioluminescent light source overlaps with the absorption spectrum of the photoluminescent material of the light guide. 6. The sight device as claimed in claim 1, wherein in the visible range at least 50%, preferably at least 70%, of the emission spectrum of the radioluminescent light source overlaps with the emission spectrum of the photoluminescent material of the light guide. 7. The sight device as claimed in claim 1, wherein the emission spectrum of the radioluminescent light source is in the green and/or yellow wavelength range. 8. The sight device as claimed in claim 1, wherein the emission spectrum of the photoluminescent material of the light guide is in the green wavelength range. 9. The sight device as claimed in claim 1, wherein the radioluminescent light source is arranged at an end side of the light guide, whereby light of the radioluminescent light source is directed through the end side into the light guide. 10. The sight device as claimed in claim 9, wherein the end side of the light guide is adhered by means of a transparent adhesive to the radioluminescent light source. 11. The sight device as claimed in claim 9, wherein the radioluminescent light source has a longitudinal extension which is perpendicular to the axis of the light guide in its end section. 12. The sight device as claimed in claim 9, wherein the end side of the light guide facing the radioluminescent light source is a polished surface. 13. The sight device as claimed in claim 1, further comprising a reverse prism, preferably a Schmidt-Pechan prism, arranged in the beam path,wherein the end side of the light guide, which faces away from the radioluminescent light source, is aligned to an in particular circular opening in a mirrored plane surface of the reverse prism (12). 14. The sight device as claimed in claim 1, wherein the end side of the light guide facing away from the radioluminescent light source is a polished surface, which preferably faces a prism for directing the light into a beam path of the sight device. 15. The sight device as claimed in claim 1, wherein the opaque coating has a white color. 16. The sight device as claimed in claim 1, wherein the radioluminescent light source and an end section of the light guide bordering the radioluminescent light source are surrounded by a housing essentially in a form-fitting manner. 17. The sight device as claimed in claim 16, wherein the housing is made in two parts, wherein preferably the two parts can be pivoted relative to one another or can be held together by means of a snap device. 18. The sight device as claimed in claim 16, wherein the housing comprises at least one opening, which leads from the outside to the coupling point between the radioluminescent light source and the light guide, in particular for introducing an adhesive. 19. The sight device as claimed in claim 16, wherein in the housing at least one screw sits in a screw thread, via which the radioluminescent light source and/or an end section of the light guide bordering the radioluminescent light source is/are clamped. 20. The sight device as claimed in claim 1, wherein the radioluminescent light source and an end section of the light guide bordering the radioluminescent light source are surrounded by an in particular T-shaped shrink tube. 21. The sight device as claimed in claim 1, wherein the radioluminescent light source together with an end section of the light guide bordering the radioluminescent light source is molded into a material. 22. The sight device as claimed in claim 1, wherein the opaque coating has a color pigmented with TiO2.
051777746	abstract	A reflection soft X-ray microscope is provided by generating soft X-ray beams, condensing the X-ray beams to strike a surface of an object at a predetermined angle, and focusing the X-ray beams reflected from the surface onto a detector, for recording an image of the surface or near surface features of the object under observation.

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