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abstract | A water-chamber working apparatus 1 according to the present invention includes a movable body that can move along a tube plate 12 of a steam generator 10, an extendable member 21 that extends and retracts in a direction in which a first coupling portion 21d approach each other and a direction in which these portions move away from each other, where the first coupling portion 21d is attached to a maintenance hatch 15 via a first joint 23a including two rotation axes intersecting with each other, and the second coupling portion 21e is attached to the movable body via a second joint 23b including two rotation axes intersecting with each other, which are different from the rotation axes of the first joint 23a. |
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claims | 1. A method of passively, safely maintaining a coolant level of a nuclear power generating facility, above a nuclear core at a preprogrammed level for an extended period of time during a facility outage in which a reactor vessel housing the nuclear core is substantially depressurized, the power generating facility comprising:a containment building that houses the reactor vessel which has an elongated axial dimension that surrounds the nuclear core in which fission reactions take place, and an open end of the reactor vessel is axially spaced from the nuclear core, with the open end sealed by a head at a reactor vessel flange;a spent fuel pool supported outside the containment building at an elevation that extends substantially above the reactor vessel, the spent fuel pool being in fluid communication with an interior of the reactor vessel through a first valve that is configured to automatically supply coolant from the spent fuel pool to the interior of the reactor vessel when a sensed level of coolant within the reactor vessel is below a given level; andan ultimate heat sink coolant reservoir whose upper level of a coolant under normal operation of the nuclear power generating facility is supported at an elevation substantially above the spent fuel pool, with a lower portion of the ultimate heat sink coolant reservoir in fluid communication with the spent fuel pool through a second valve whose operation is automatically controlled by a level of coolant in the spent fuel pool to maintain the coolant in the spent fuel pool at approximately a preselected level;the method including the steps of:sensing coolant level within the reactor vessel above the nuclear core from a gauge on a branch coolant line connected to the reactor vessel, having an output indicative of a coolant level above the nuclear core;automatically controlling the first valve to drain coolant from the spent fuel pool into the reactor vessel when the sensing step identifies the coolant level is at the given level to maintain the coolant level within the reactor vessel at the preprogrammed level above the nuclear core; andautomatically controlling the second valve to drain coolant from the ultimate heat sink coolant reservoir into the spent fuel pool to maintain the coolant in the spent fuel pool at approximately the preselected level. 2. The method of claim 1 in which the nuclear power generating facility has a station blackout including the steps of:opening the first and second valves; andflooding the reactor vessel. 3. The method of claim 1 wherein the preprogrammed level is approximately at the reactor vessel flange. 4. The method of claim 1 wherein the nuclear power generating facility includes a refueling cavity supported above the reactor vessel flange and the reactor vessel head has been removed, the gauge controls the level of coolant above the nuclear core within the refueling cavity. 5. The method of claim 4 wherein the nuclear power generating facility includes a refueling canal establishing a fluid communication path between an inside of the refueling cavity at an elevation above the reactor vessel flange, and the spent fuel pool, through which a fuel assembly can pass, and means for isolating the fluid communication path from the inside of the refueling cavity, including the steps of:opening the means for isolating the fluid communication path; andcontrolling a level of the coolant within the refueling cavity through the fluid communication path. 6. The method of claim 5 in which the nuclear power generating facility has a station blackout including the steps of:opening the first valve; andflooding the containment building. |
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040452840 | abstract | A nuclear reactor fuel containment safety structure is disclosed herein and is shown to include an atomic reactor fuel shield with a fuel containment chamber and exhaust passage means, and a deactivating containment base attached beneath the fuel reactor shield and having exhaust passages, manifold, and fluxing and control material and vessels. |
abstract | Method and apparatus for containing, transporting, and storing or disposing of radioactive machinery, including decommissioned nuclear reactor pressure vessels. An improved, economically-produced container is provided which allows easier handling and packaging of machinery within plants where the machinery has been installed, and which provides improved shock absorption and attenuation characteristics, especially when packaging is complete. A reactor pressure vessel or similar item is disconnected from the remainder of the plant and external fittings are trimmed as close to flush with item""s exterior as practicable. A storage and containment canister, optionally cut into at least two sections to ease handling and packaging, is placed nearby. The pressure vessel head or any other low-radioactive items are removed, and insulation and other items removed from the outside of the item are placed inside the item""s body. The item body is placed into a lower section of the canister and sealed, and detached canister sections are reattached. Gaps between the canister and the RPV body and preferably the interior of the RPV body are filled with grout or low denisity cellular concrete. The canister is closed, and the pressure vessel head or other removed portions are secured to the outside of the canister. Optionally the canister exterior is sealed with a metalizing spray. The complete package is transported for storage or disposal. |
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description | This application is related to commonly-owned U.S. patent application Ser. No. 12/022,765 filed Jan. 30, 2008 and entitled “Dynamic Real-Time Power System Monitoring.” Embodiments of the invention relate generally to electric power transmission systems and, more particularly, to monitoring of electric power transmission systems using a time synchronized programmable discrete interval binary pseudorandom sequence signal injection system. Monitoring the state of the electric power grid for operating anomalies, as well as equipment degradation and failures, is important for the reliable supply of electric energy. Current methods involve monitoring parameters such as voltage, current and relative phase angles of the power system as it operates. These actual values are then applied to an electrical model for analysis. These conventional monitoring methods typically require a steady state operational environment for power systems analysis. Actual power system measurements, from a subset of all key data points in the power system, are typically captured using a Supervisory Control and Data Acquisition (SCADA) system. Each substation connected to the power grid is equipped with several potential and current transformers to measure voltage, current, and electric power flow on each line and bus. The real-time voltage and current data is transmitted from each substation to a central computer through a remote terminal unit. These acquired readings from throughout the power system and are then processed by a state estimator algorithm to determine a complete set of the most likely values for all key points in a model of the power system. System security applications are run on these models to assess the ability of the power system to recover from various possible disturbances. This security contingency analysis attempts to determine if the power system will return to an equilibrium state or become unstable after selected system disturbances. One characteristic of power system modeling for conventional security analysis is that the system topology and component parameters must be correct for the results to be meaningful. Inaccuracies in conventional security analysis can occur if data sampled from the actual power system are applied to a model that fails to consider a change in system topology, e.g., an open transmission line in its connectivity topology. Furthermore, inaccuracies in the models used for various power system components, such as transmission lines and power plants, can lead to an inaccurate security analysis. The embodiments of the time synchronized programmable discrete interval binary pseudorandom sequence signal injection system described herein are used for determining real-time dynamic impulse and frequency response characteristics of the transmission grid and its components. These characteristics are aggregated to provide real-time dynamic monitoring of the state of the power grid and its components. Embodiments of the time synchronized programmable discrete interval binary pseudorandom sequence signal injection system provide a system and method for creating and injecting synchronized signals into the electric power grid for the purpose of determining real-time dynamic characteristics of the grid and its components. Injected signals are time synchronized with data capture devices at other locations in the power system by using a time sync system such as Global Positioning System (GPS) time synchronization clock units. Injected sequences are sets of binary pseudorandom sequences (PRS). Sets of binary pseudorandom sequences can be generated mathematically using software algorithms or with hardware logic through the use of maximum length Linear Feedback Shift Registers (LFSR). Injected sequence parameters include, but are not limited to, bit rate, sequence length, sequence bit order, and injected signal magnitude. Selection of the various injected sequence parameters is through a manual control interface on the device itself or remotely over a communications link. The sequence may be injected directly on to the power system through an interface device such as a capacitor coupled voltage transformer (CCVT), or inductively through a clamp-on inductive coupling device. In an exemplary embodiment, a synchronized pseudorandom sequence injector is provided for injecting a plurality of pseudorandom signals at selected locations in a power system having a plurality of locations forming a transmission and distribution grid. A synchronization pulse generator generates an accurate reference clocking signal. A pseudorandom clocking and sequence generator receives the clocking signal and generates a string of pseudorandom sequences. A binary drive control creates a tri-state voltage output from a logic level output of the pseudorandom clocking and sequence generator. A signal conditioning interface processes the voltage output to attenuate any protection related carrier signals from a pseudorandom signal injection point at a selected location. The following description is provided as an enabling teaching of the invention and its best, currently known embodiments. Those skilled in the relevant art will recognize that many changes can be made to the embodiments described, while still obtaining the beneficial results. It will also be apparent that some of the desired benefits of the embodiments described can be obtained by selecting some of the features of the embodiments without utilizing other features. Accordingly, those who work in the art will recognize that many modifications and adaptations to the embodiments described are possible and may even be desirable in certain circumstances, and are a part of the invention. Thus, the following description is provided as illustrative of the principles of the embodiments of the invention and not in limitation thereof, since the scope of the invention is defined by the claims. A new approach to monitoring the power system is based on processing the synchronized responses to injected pseudorandom signals at various locations throughout the grid. Real-time analysis of the response data can be used to provide a characterization of the current state of the power grid and its various components. A complete description of a method and system for real-time monitoring of a power system is provided in patent application Ser. No. 12/022,765, now U.S. Pat. No. 7,848,897, which is incorporated by reference in its entirety herein. FIG. 1 illustrates an exemplary embodiment of the overall real-time dynamic transmission monitoring system incorporating the time synchronized programmable discrete interval binary pseudorandom sequence signal injection system. As described in more detail in related application Ser. No. 12/022,765, a fundamental concept of the independent dynamic transmission monitoring system is the system impulse response function, generally designated h(t). The impulse response of a system is the output of the system when it is presented with a very brief yet large input, i.e., an impulse. A given well-behaved system is completely described by its impulse response function and useful information can be derived directly from the time domain h(t) function. The impulse response function can be used to determine the output of a system to any given input function through the use of the convolution integral. By applying a transform function to the time-domain impulse response function h(t), the system frequency response can be determined. The Fourier transform H(ω) thus becomes the frequency domain description of the system and can be used to determine the dynamic characteristics of the system. In signal processing the cross-correlation between two signals can be described as the measure of the similarity of the two signals. A single cross-correlation value is calculated for each value of the time shift between the two signals. Autocorrelation is defined as the cross-correlation of a time domain signal with itself. For a “well behaved” system, the impulse response can be determined using cross-correlation techniques and random input signals. The impulse response h(t) can be determined by cross-correlating a random input signal ni (t) with the resulting system output y(t). The requirements are: (1) the random input signal generated by pseudorandom signal generator must be uncorrelated with the other inputs, and (2) the autocorrelation of the random input signal must be an impulse. A cross-correlation algorithm for performing impulse response calculations using pseudorandom input signals has been described in U.S. Pat. No. 3,718,813, entitled “Technique for Correlation Method of Determining System Impulse Response.” This patent is incorporated by reference herein. This algorithm calculates a system impulse response waveform during each repeated sequence of the pseudorandom noise signal. Use of this algorithm can potentially reduce the complexity and cost of a field-implemented cross-correlator to that of an inexpensive microcontroller. The real-time cross-correlation calculation technique can be used between selected locations throughout the electric grid to calculate multiple simultaneous impulse response data sets and therefore provide an overall dynamic real-time image of the instantaneous status of the entire power system. An overall status of the “state of connectedness” of the power grid can be monitored in real-time and, by using visualization techniques, this can be used to provide operators with graphic indicators of where and how intense “operating weaknesses” are in the system. A central host system based on a high performance computing (HPC) cluster architecture collects and processes the array of continuous impulse response calculations and creates numeric and visual representations of the dynamic state of the transmission system. FIG. 1 illustrates an exemplary dynamic transmission monitoring system using the cross-correlation method. The synchronized pseudorandom signal (PRS) injectors and the synchronized impulse response and parameter calculators are located in substations and inter-utility tie points throughout the power system. They communicate with the monitoring center through the Internet using HyperText Transfer Protocol (HTTP). The architecture of the high performance computing cluster monitoring center depicted in FIG. 1 for an exemplary embodiment includes various compute nodes that support the parallel computational and input/output (I/O) functions described in the following paragraphs. PRN nodes support the functions of starting, stopping and adjusting the parameters of the synchronized pseudorandom signal injectors located throughout the power system. H(t) Nodes support the functions of starting, stopping and acquiring the cross-correlation data from the synchronized impulse response and parameter calculators located throughout the power system. MMI Nodes respond to requests from system operators, engineers, planners and coordinators at MMI consoles for real-time dynamic power system configuration information by generating numeric and graphical displays of the state of the power system. VIS Nodes generate a high resolution multi-screen video wall depicting three dimensional (3D) visualizations of the real-time dynamic power system state based on the collected impulse response data from locations throughout the power system. The video wall can be controlled by requests from the various MMI consoles. Compute Nodes perform calculations and analysis on the collected impulse response data from H(t) nodes to produce numeric and graphical information for use by MMI nodes and VIS nodes. Host Node performs cluster management, initializing, starting, stopping and altering cluster processes on the various compute nodes through the cluster console. The host node also incorporates at least one data store hard disk for cluster initialization and archival information storage. The low latency network fabric of the depicted HPC Cluster is based on Gigabit Ethernet. Each compute node connects to a Gigabit Ethernet switch and inter-node communications is accomplished using Message Passing Interface (MPI) libraries. MPI is the industry-standard Message Passing Interface. The architecture depicted in FIG. 1 is identical to the architecture used by the vast majority of the fastest, most powerful supercomputers currently available. The depicted configuration uses the Linux operating system (OS), cluster technology, Gigabit Ethernet switches, and commodity off-the-shelf CPUs. Other high performance architectures also can be used to perform the functionality of the components shown in FIG. 1. The depicted HPC cluster uses diskless compute nodes that boot using PXE (Preboot eXecution Environment). In this environment, the compute nodes are booted from a TFTP server running on the Host Node. The PXE boot process loads the operating system and initial applications into the compute node's random access memory and starts initial execution of the node processes. The firewall router(s) is used to interface the PRN nodes and the H(t) nodes to the Internet where they communicate over HTTP to web servers in the substations located throughout the power system. The following paragraphs describe the various users at the MMI consoles of the monitoring system depicted in FIG. 1. The users include system operators, system engineers, and system planners. System operators normally operate the power system. System operators continually monitor the normal power system operating parameters and execute routine system operating procedures such as pre-arranged switching. When system anomalies occur, the system operators take pre-established remedial action to restore the power system to its normal operating state. System engineers have engineering knowledge of the power system and review changes to the power system configuration for stability and security. System engineers develop the operating procedures that system operators follow. System planners use the historical power system data and customer needs to plan future additions and modifications to the power system. The time synchronized programmable discrete interval binary pseudorandom sequence signal injection system is an integral part of the monitoring system for the electric power transmission system and it components. The function of the signal injection system is to generate pseudorandom sequences that are synchronized with data capture elements around the transmission grid. The captured data can be cross-correlated with the injected pseudorandom sequence (PRS) to produce the power system impulse response, h(t), between the injection point and the data capture location. FIG. 2 illustrates the basic underlying theory of using injected pseudorandom signal and cross-correlation techniques to determine the real-time dynamic characteristics of a system. As shown in FIG. 2, the time synchronization of a pseudorandom signal generated by pseudorandom sequence generator in the signal injector with the data analysis (e.g., cross-correlation) calculations in the data capture element generates the impulse response h(t) at a separate location in the power system from the location where the signal is being injected. This provides dynamic power system information without relying on an accurate mathematical model of the grid. The actual real world system response information between the two locations on the grid is obtained dynamically in real-time using this approach. Multiple synchronized data samples can be captured around the transmission system in parallel and cross-correlated with the injected sequence to produce a set of impulse responses from the PRS injection point to the various data capture locations on the grid. Binary pseudorandom sequences can be generated using a linear feedback shift register (LFSR) with appropriate odd parity feedback from certain elements or stages of the shift register, referred to as taps. Only specific sets of taps produce what is referred to as a maximum length shift register sequence, or a pseudorandom sequence. The discrete interval of these binary sequences is determined by the frequency of the shift register clocking. Although these sequences are absolutely deterministic, they possess statistical randomness properties that make them suitable for determining impulse responses using the cross-correlation method. FIG. 5, discussed below, illustrates a variable length shift register as a component of the PRS clocking and sequence generator in an embodiment of the invention. Binary pseudorandom sequences generated from an LFSR with n stages and appropriate feedback taps will produce a sequence that is 2n−1 bits in length. Typically there are many unique binary pseudorandom sequences that can be generated for a given LFSR of length n based on selecting appropriate feedback taps. Tap tables yielding pseudorandom sequences have been discovered and are widely published. A useful characteristic of different binary pseudorandom sequences, even different sequences of the same length n, is that they are uncorrelated with each other. As a result of this characteristic, two or more different sequences can be injected into the electric grid at different locations and the synchronized captured data at another location on the grid can be cross-correlated with each of the different sequences to produce the unique impulse responses from the injection locations to the data capture point. FIG. 5 illustrates an exemplary pseudorandom sequence generation subsystem depicting an adjustable bit rate clock, adjustable length Linear Feedback Shift Register (LFSR) and adjustable feedback taps. A switch selectable divide by n counter circuit reduces the crystal oscillator frequency to the desired PRS bit rate and is used to clock the shift register. An “exclusive OR operation” (XOR, odd parity) on selected outputs from the various stages of the shift register provides the feedback signal to the beginning stage of the shift register. Only certain selected outputs, referred to as “Taps,” will produce maximum length LFSR sequences that are pseudorandom. In the embodiment shown in FIG. 5 the initial stages of the shift register are implemented individually with selectable serial flow control data switches to adjust the configuration. These can be electronic data flow switches or simply configurable hardware jumpers on a printed circuit board. In this example, by directing the XOR feedback signal directly into the 8 bit shift register element and ignoring the four external elements, a LFSR of length 28−1 will be configured. To configure a LFSR of length 212−1, referred to as a PN12 configuration, the feedback signal is applied to the left most shift register element and the data flow is directed from the output of each of the external elements to the input of the next, creating a shift register with 12 total elements, i.e., four individual external and eight combined in a single unit. The embodiment shown in FIG. 5 is designed to produce sequences PN8, PN9, PN10, PN11, and PN12 by selecting the appropriate serial data flow configurations. FIG. 3 illustrates an exemplary block diagram of the basic components of the time synchronized programmable discrete interval binary pseudorandom sequence signal injection system when used with a coupling capacitor voltage transformer (CCVT) to interface with the electric power transmission grid. The pseudorandom sequence signal injection system includes a GPS clock device 300 that generates an accurate synchronization pulse to the PRS signal injector at the same time similar GPS clock devices initiate the data capture events at other locations on the transmission grid. The GPS clock device initiates the generation of a string of pseudorandom sequences created by the PRS clocking and sequence generator 310. A binary drive control 320 creates a tri-state voltage output from the logic level output of the PRS clocking and sequence generator. This voltage signal is then processed by a signal conditioning unit 330 to attenuate any protection related carrier signals and block any 60 Hz and direct current components from the CCVT. Electric utilities typically use devices, referred to as coupling capacitor voltage transformers (CCVT), to inject one or more carrier frequencies on to a given transmission line as part of the protection scheme. These devices, where they are implemented, offer an ideal way to inject a binary pseudorandom sequence on to the electric power transmission grid. FIGS. 8A-8B illustrate an exemplary sample circuit diagram and a sketch of a coupling capacitor voltage transformer (CCVT) for use on the electric power transmission grid. A CCVT is used in the power industry on high voltage transmission systems as a voltage divider to provide a lower operating voltage for protective relaying and metering equipment. CCVTs also provide a means of inserting and receiving carrier signals over the high voltage transmission lines. FIG. 8A illustrates a typical schematic representation of a CCVT showing the stack of capacitive elements comprising what is referred to as C1 and the lower element C2. Values of the capacitive elements are chosen such that voltage that appears across C2 is a small fraction of the high voltage appearing on the transmission line relative to earth ground. A drain coil in parallel with a spark gap and a grounding switch connects the capacitor stack to earth ground. The drain coil offers a very low impedance to the power line frequency (typically 50 to 60 Hz) and a very high impedance to the carrier signal frequencies (typically 30 to 300 KHz). Conditioned PRS signals injected at the drain coil therefore travel to the transmission line through the capacitor stack. FIG. 8B illustrates the typical form of a CCVT with its insulated capacitor stack mounted on a terminal box containing the drain coil and other carrier tuning equipment. To facilitate deployment of PRS injectors and data capture devices throughout the power grid, especially outside of substations, a quick connect/disconnect clamp-on device is desirable. FIG. 7 illustrates an exemplary inductively coupled wireless clamp-on PRS injector and data capture device for use on the electric power transmission grid. The device can be installed and removed without de-energizing the transmission line by placing it over the conductor and securing the movable latch that also completes the inductive circuit. Photovoltaic cells configured around the top of the base provide charging current for the re-chargeable batteries used to power the unit. The unit is controlled by a microprocessor that interfaces with the hardwired PRS injector unit transmission line data capture elements and the wireless communications transceiver. For this embodiment, the wireless transceiver uses 3G or 4G broadband technology for Internet connectivity. Communications are encrypted for data security. The unit also contains a GPS receiver for device location determination and time synchronization of injected sequences and data capture. FIG. 4 illustrates the deployment of both CCVTs and wireless clamp-on devices throughout the electric power transmission grid for synchronized PRS injection and data capture. Injector parameters can be set using a manual control panel as depicted in FIG. 9 or though a wireless 3G or 4G broadband Internet connection with the host system as illustrated in the system architecture of FIG. 1. A manual setup of the operating parameters is performed locally via the exemplary control panel shown in FIG. 9. The four sequence length selection switches labeled PN9-PN12 correspond to shift registers of 9-12 stages, respectively. For example, PN9 represents a shift register having nine stages which will cause the PRS signal injector to inject a signal sequence of 29−1 bits in length. The frequency selection switches below the PN9-PN12 switches enable the user to select the PRS bit clock frequency. The sequence control push buttons below the frequency selection switches are used to arm, start, and stop the PRS injection sequence. The power connector is located to the right of the sequence control push buttons. The left side of the control panel depicts the BNC connectors. Tap selection jumpers are located between the BNC connectors and the frequency selection switches on the control panel. While the use of a microprocessor-based PRS injector offers wireless remote control of the injector parameters, there is always the threat of a cyber security breach. Thus, the most secure implementation is based on a fixed hardwired PRS injector as illustrated in FIG. 10. The pseudorandom sequence bit rate timing begins with an oscillator that generates a binary clock signal. In the exemplary embodiment of FIG. 10, the oscillator is a crystal-based 40 MHz printed circuit board unit. The 40 MHz signal is divided by two using a D flip-flop logic element labeled U2a to provide a 50% duty cycle for the clock timing. This 20 MHz clock signal then clocks an eight bit counter, made up of two four bit synchronous counters labeled U7 and U8 in series. The output of this eight bit counter is the pseudorandom sequence bit rate clock. The eight bit counter is preset to a specific count based on the setting of a set of dip switches labeled SW1 and SW2. The counter counts from the specific count down to zero at which time the counter is reset to the count given by the dip switch setting. Therefore, the eight bit dip switch setting determines the pseudorandom sequence bit rate. The LFSR is made up of four D flip-flop logic elements labeled U3a, U3b, U4a, and U4b, plus an eight bit shift register logic element labeled U6. The sixteen pin header labeled J1 provides a means to select specific feedback taps from elements of the LFSR that are then applied to the XOR gates labeled U14. Jumper J2 provides a means to select a two-tap or a four-tap configuration to generate the odd parity XOR feedback signal. Jumpers labeled J3, J4, J5 and J6 provide for configuration of a PN9, PN10, PN11 or a PN12 length LFSR. Two eight-input NAND logic gates labeled U10 and U11 along with NOR logic gate labeled U13a detect the state in which all elements of the LFSR are a logic “1” and through connector labeled H4 provide a one-bit time logic pulse at the beginning of each pseudorandom sequence. The STOP push button on the control panel resets both D flip-flops labeled U5a and U4b to a logic “0” and energizes the red STOP light-emitting diode (LED). This also holds a preset on D flip-flops labeled U3a, U3b, U4a, and U4b allowing only logic “1's” to be clocked out of the LFSR. The ARM push button on the control panel sets the D flip-flop U5b to a logic “1” and energizes the amber ARM LED. If D flip-flop U5b is in the ARM state (logic “1”), then D flip-flop U5a can be set to a logic “1” by either the START push button on the control panel or a GPS clock pulse. When D flip-flop U5a is set to a logic “1” the green RUN LED is energized and the LFSR will clock through pseudorandom sequences if appropriate taps have been selected. FIG. 6 illustrates processing logic for the pseudorandom sequence generation subsystem in an exemplary embodiment. Processing starts in step 600. In step 604, the following values are set: (1) output voltage level, (2) sequence count (n), (3) PRS length, and (4) PRS bit clock rate (e.g., via the PRS bit clock frequency switches of FIG. 9). In step 608, the run/stop function is set to stop. Next, in decision step 612, a determination is made if the ARM function is on (ARM=1). If it is, then in decision step 616 a determination is made if the GPS clock function is on (GPS clock=1). If the GPS clock function is on, then the run/stop function is set to run in step 620. A determination is then made in decision step 624 if the sequence has finished. If the sequence has finished, a determination is made in decision step 628 if the sequence count is equal to the set sequence count. If the sequence count equals the set sequence count, the run/stop function is set to stop as indicated in step 632. Processing logic returns to the start step in connector step 636. The corresponding structures, materials, acts, and equivalents of all means plus function elements in any claims below are intended to include any structure, material, or acts for performing the function in combination with other claim elements as specifically claimed. Those skilled in the art will appreciate that many modifications to the exemplary embodiments are possible without departing from the scope of the present invention. In addition, it is possible to use some of the features of the embodiments described without the corresponding use of other features. Accordingly, the foregoing description of the exemplary embodiments is provided for the purpose of illustrating the principles of the invention, and not in limitation thereof, since the scope of the invention is defined solely by the appended claims. |
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claims | 1. A support grid for a nuclear fuel assembly, said nuclear fuel assembly including a generally cylindrical fuel rod with a diameter, said support grid comprising:a plurality of generally tubular frame members defining frame assembly including a plurality of generally cylindrical cells, each said cell including at least one sidewall;at least one generally cylindrical tubular member including a cell contact portion with a greater diameter and at least one helical fuel rod contact portion with a lesser diameter, said cell contact portion and said fuel rod contact portion joined by a transition portion, said greater diameter being generally equivalent to said cell radius, and said lesser diameter being generally equivalent to said fuel rod diameter such that a fuel rod disposed in said tubular member would engage said lesser diameter; andeach of said at least one tubular member disposed in one cell of said plurality of generally circular cells so that said cell contact portion engages said cell sidewall. 2. The support grid of claim 1 wherein each said at least one tubular member has a wall with a uniform thickness. 3. The support grid of claim 1 wherein each said at least one helical fuel rod contact portion extends 360 degrees about said tubular member. 4. The support grid of claim 1 wherein each said at least one helical fuel rod contact portion includes two helical fuel rod contact portions. 5. The support grid of claim 4 wherein each said helical fuel rod contact portion extends 180 degrees about said tubular member. 6. The support grid of claim 4 wherein each said helical fuel rod contact portion extends 90 degrees about said tubular member. 7. The support grid of claim 1 wherein each said transition portion is a generally smooth curve. 8. The support grid of claim 1 wherein:said cell contact portion includes a platform; andsaid transition portion is a sharp curve. 9. The support grid of claim 8 wherein each said fuel rod contact portion includes a concave platform. 10. The support grid of claim 1 wherein:said fuel rod contact portion includes a concave platform; andsaid transition portion is a sharp curve. 11. The support grid of claim 1 wherein:said cell contact portion includes a platform; andsaid transition portion is generally flat with angled ends. 12. The support grid of claim 11 wherein each said fuel rod contact portion includes a concave platform. 13. The support grid of claim 1 wherein:said fuel rod contact portion includes a concave platform; andsaid transition portion is generally flat with angled ends. 14. The support grid of claim 1 wherein:said frame assembly includes a plurality of generally tubular frame members defining said generally cylindrical cells; andeach said tubular frame members coupled to each other at 90 degree intervals about the perimeter of each tubular frame member. 15. The support grid of claim 1 wherein each said at least one helical fuel rod contact portion has a variable pitch. 16. The support grid of claim 1 wherein each said at least one helical fuel rod contact portion has a constant pitch. 17. The support grid of claim 1 wherein:said tubular member has a cylinder axis and at least a first axial portion and a second axial portion;said at least one helical fuel rod contact portion extending over both said first axial portion and said second axial portion; andwherein said at least one helical fuel rod contact portion on said first axial portion has a first pitch and said at least one helical fuel rod contact portion on said second axial portion has a second pitch. 18. The support grid of claim 1 wherein:each said each said at least one helical fuel rod contact portion may have a clockwise twist or a counterclockwise twist;at least one tubular member including a helical fuel rod contact portion with a clockwise twist; andat least one tubular member including a helical fuel rod contact portion with a counterclockwise twist. 19. The support grid of claim 1 wherein each said at least one helical fuel rod contact portion includes four helical fuel rod contact portions. 20. The support grid of claim 7 wherein each said helical fuel rod contact portion extends 90 degrees about said tubular member. |
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claims | 1. A system for creating an image of an object, said object containing a first chemical element having a K EDGE, FIRST CHEMICAL ELEMENT , said system comprising: an X-ray source configured to emit an X-ray beam along a first path through said object; means for moving said beam onto successive paths through said object; a first filter and a second filter, said first filter comprising a chemical element having a K EDGE greater than K EDGE, FIRST CHEMICAL ELEMENT and said second filter comprising a chemical element having a K EDGE less than K EDGE, FIRST CHEMICAL ELEMENT ; a means for successively positioning said filters on each said path to filter said beam and produce a first filtered radiation signal and a second filtered radiation signal for each said path through said object; a detector for receiving said first and second radiation signals after said signals have passed through said object and producing a first and second electrical signal for each said path, said electrical signals produced by said detector being proportional to the intensity of the radiation received by said detector; and a processor connected to said detector to generate an image signal for each said path through said object wherein said image signal is proportional to a difference between said first and second electrical signals, said processor configured to produce an image of said object from said image signals. 2. A system as recited in claim 1 wherein said means for successively positioning said filters comprises: claim 1 a wheel defining an axis and formed with a pair of holes, each said hole for accommodating a said filter; and means for selectively rotating said wheel about said axis. 3. A system as recited in claim 1 wherein said means for selectively rotating said wheel about said axis is a motor and said motor and wheel are mounted on said X-ray source for movement therewith. claim 1 4. A system for creating an image of an object that contains a contrast agent, said system comprising: a means for directing a spectrum of electromagnetic radiation onto a plurality of paths, each said path extending through said object; a first filter and a second filter, each said filter comprising at least one chemical element having a K EDGE within said spectrum of said electromagnetic radiation; a means for successively interposing each said filter on each said path to create a first filtered radiation signal and a second filtered radiation signal for each said path; a means for receiving said filtered radiation signals and producing an electrical signal for each said filtered radiation signal received, each said electrical signal being proportional to the intensity of radiation received; and a processor for operation on said electrical signals to produce an image of said object. 5. A system as recited in claim 4 wherein said interposing means is positioned to successively interpose each said filter on each said path to contact said radiation before said radiation passes through said object. claim 4 6. A system as recited in claim 4 wherein said means for successively interposing each said filter on each said path comprises: claim 4 a wheel defining an axis and formed with a plurality of holes, each said hole for accommodating a said filter; and means for selectively rotating said wheel about said axis. 7. A system as recited in claim 4 wherein said contrast agent comprises a chemical element having a K EDGE, CONTRAST AGENT and said first filter comprises a chemical element having a K EDGE greater than K EDGE, CONTRAST AGENT , and said second filter comprises a chemical element having a K EDGE less than K EDGE, CONTRAST AGENT . claim 4 8. A system as recited in claim 7 wherein said contrast agent comprises the chemical element Iodine, said first filter comprises a layer of the chemical element Cs at an approximate thickness of 290 xcexcm and said second filter comprises a layer of the chemical element Te at an approximate thickness of 102 xcexcm. claim 7 9. A system as recited in claim 7 wherein said contrast agent comprises the chemical element Xe, said first filter comprises a layer of the chemical element Ba at an approximate thickness of 180 xcexcm and said second filter comprises a layer of the chemical element Te at an approximate thickness of 123 xcexcm. claim 7 10. A system as recited in claim 7 wherein said contrast agent comprises the chemical element Cs, said first filter comprises a layer of the chemical element Ba at an approximate thickness of 180 xcexcm and said second filter comprises a layer of the chemical element I at an approximate thickness of 148 xcexcm. claim 7 11. A system as recited in claim 7 wherein said contrast agent comprises the chemical element Ba, said first filter comprises a layer of the chemical element La at an approximate thickness of 105 xcexcm and said second filter comprises a layer of the chemical element Cs at an approximate thickness of 380 xcexcm. claim 7 12. A system as recited in claim 7 wherein said contrast agent comprises the chemical element Sm, said first filter comprises a layer of the chemical element Eu at an approximate thickness of 136 xcexcm and said second filter comprises a layer of the chemical element Nd at an approximate thickness of 117 xcexcm. claim 7 13. A system as recited in claim 7 wherein said contrast agent comprises the chemical element Eu, said first filter comprises a layer of the chemical element Gd at an approximate thickness of 120 xcexcm and said second filter comprises a layer of the chemical element Sm at an approximate thickness of 138 xcexcm. claim 7 14. A system as recited in claim 7 wherein said contrast agent comprises the chemical element Gd, said first filter comprises a layer of the chemical element Tb at an approximate thickness of 140 xcexcm and said second filter comprises a layer of the chemical element Eu at an approximate thickness of 236 xcexcm. claim 7 15. A system as recited in claim 7 wherein said contrast agent comprises the chemical element Tb, said first filter comprises a layer of the chemical element Dy at an approximate thickness of 130 xcexcm and said second filter comprises a layer of the chemical element Gd at an approximate thickness of 151 xcexcm. claim 7 16. A system as recited in claim 7 wherein said contrast agent comprises the chemical element Dy, said first filter comprises a layer of the chemical element Ho at an approximate thickness of 130 xcexcm and said second filter comprises a layer of the chemical element Tb at an approximate thickness of 149 xcexcm. claim 7 17. A system as recited in claim 7 wherein said contrast agent comprises the chemical element Ho, said first filter comprises a layer of the chemical element Er at an approximate thickness of 130 xcexcm and said second filter comprises a layer of the chemical element Dy at an approximate thickness of 149 xcexcm. claim 7 18. A system as recited in claim 7 wherein said contrast agent comprises the chemical element Er, said first filter comprises a layer of the chemical element Tm at an approximate thickness of 135 xcexcm and said second filter comprises a layer of the chemical element Ho at an approximate thickness of 156 xcexcm. claim 7 19. A system as recited in claim 7 wherein said contrast agent comprises the chemical element Lu, said first filter comprises a layer of the chemical element Hf at an approximate thickness of 110 xcexcm and said second filter comprises a layer of the chemical element Yb at an approximate thickness of 235 xcexcm. claim 7 20. A system as recited in claim 7 wherein said contrast agent comprises the chemical element Hf, said first filter comprises a layer of the chemical element Ta at an approximate thickness of 97 xcexcm and said second filter comprises a layer of the chemical element Lu at an approximate thickness of 175 xcexcm. claim 7 21. A system as recited in claim 7 wherein said contrast agent comprises the chemical element Ta, said first filter comprises a layer of the chemical element W at an approximate thickness of 84 xcexcm and said second filter comprises a layer of the chemical element Hf at an approximate thickness of 131 xcexcm. claim 7 22. A system as recited in claim 7 wherein said contrast agent comprises the chemical element W, said first filter comprises a layer of the chemical element Re at an approximate thickness of 80 xcexcm and said second filter comprises a layer of the chemical element Ta at an approximate thickness of 108 xcexcm. claim 7 23. A system as recited in claim 7 wherein said contrast agent comprises the chemical element Re, said first filter comprises a layer of the chemical element Os at an approximate thickness of 80 xcexcm and said second filter comprises a layer of the chemical element W at an approximate thickness of 100 xcexcm. claim 7 24. A system as recited in claim 7 wherein said contrast agent comprises the chemical element Os, said first filter comprises a layer of the chemical element Ir at an approximate thickness of 80 xcexcm and said second filter comprises a layer of the chemical element Re at an approximate thickness of 90 xcexcm. claim 7 25. A system as recited in claim 7 wherein said contrast agent comprises the chemical element Ir, said first filter comprises a layer of the chemical element Pt at an approximate thickness of 90 xcexcm and said second filter comprises a layer of the chemical element Os at an approximate thickness of 92 xcexcm. claim 7 26. A system as recited in claim 7 wherein said contrast agent comprises the chemical element Bi, said first filter comprises a layer of the chemical element Th at an approximate thickness of 170 xcexcm and said second filter comprises a layer of the chemical element Pb at an approximate thickness of 230 xcexcm. claim 7 27. A method for creating an image comprising the steps of: providing an object; directing a spectrum of electromagnetic radiation onto a plurality of paths, each said path extending through said object; providing a first filter and a second filter, each said filter comprising at least one chemical element having a K EDGE within said spectrum of said electromagnetic radiation; successively interposing each said filter on each said path to create a first filtered radiation signal and a second filtered radiation signal for each said path; receiving said filtered radiation signals for each said path and producing an electrical signal for each said filtered radiation signal received, each said electrical signal being proportional to the intensity of radiation received; and processing said electrical signals to produce an image of said object. 28. A method as recited in claim 27 further comprising the step of administering a contrast agent into said object. claim 27 29. A method as recited in claim 27 wherein said object is a human body. claim 27 30. A system for creating an image of an object, said object containing a first chemical element having a K EDGE, FIRST CHEMICAL ELEMENT , said system comprising: a means for directing X-ray radiation through said object; a plurality of filter pairs with each said filter pair having a first filter comprising a chemical element having a K EDGE greater than K EDGE, FIRST CHEMICAL ELEMENT and a said second filter comprising a chemical element having a K EDGE less than K EDGE, FIRST CHEMICAL ELEMENT , each said filter pair positioned to filter said radiation and produce a pair of filtered radiation signals; a plurality of detector pairs, each said detector pair for receiving one said pair of filtered radiation signals from one said filter pair and producing a pair of electrical signals in response, with each said electrical signal being proportional to the intensity of radiation received by a said detector; and a processor connected to said detector pairs to generate an image signal for each said detector pair, wherein said image signal is proportional to a difference between electrical signals in each said pair of electrical signals to produce an image of said object from said image signals. 31. A system as recited in claim 30 wherein said first and second filters of each said filter pair are positioned adjacent to each other. claim 30 32. A system as recited in claim 30 wherein said plurality of filter pairs is arranged as a planar array. claim 30 33. A system as recited in claim 32 wherein said plurality of filter pairs is arranged to create a two-dimensional, alternating pattern of first filters and second filters. claim 32 34. A system as recited in claim 30 wherein said object is a human body. claim 30 35. A system as recited in claim 30 further wherein said directing means is an X-ray source and further comprising a means for moving said X-ray source, said plurality of filter pairs and said plurality of detector pairs relative to said object during imaging of said object. claim 30 |
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claims | 1. An external reactor vessel cooling and electric power generation system, comprising:a reactor vessel;a steam generator disposed inside the reactor vessel and receiving water from a feed water system to produce steam;an external reactor vessel cooling section formed to enclose at least part of the reactor vessel so as to cool heat discharged from the reactor vessel;a power production section including a small turbine and a small generator to generate electric power using a fluid that receives heat from the external reactor vessel cooling section;a condensation heat exchange section to perform a heat exchange of the fluid discharged after operating the small turbine, and condense the fluid to generate condensed water; anda condensed water storage section to collect therein the condensed water generated in the condensation heat exchange section,wherein the fluid receiving the heat from the reactor vessel is circulated, andwherein the steam generator is connected to a large turbine and a large generator to generate electric power,wherein the electric power produced by the power production section has a capacity of less than 1% compared to the electric power produced by the large turbine and the large generator, andwherein the power production section is operated during a normal operation of a nuclear power plant and during an accident of the nuclear power plant to produce electric power. 2. The system of claim 1, wherein the condensed water in the condensed water storage section is circulated through the external reactor vessel cooling section, the power production section, and the condensation heat exchange section, and the condensed water is phase-changed into gas by the heat received from the reactor vessel. 3. The system of claim 1, further comprising:an evaporation section connected to the external reactor vessel cooling section to cause a heat exchange between a fluid inside the external reactor vessel cooling section and the condensed water of the condensed water storage section,wherein the system further comprises:a first circulation part defined between the external reactor vessel cooling section and the evaporation section such that the fluid from the external reactor vessel cooling section flows therealong; anda second circulation part provided sequentially along the evaporation section, the power production section, the condensation heat exchange section, and the condensed water storage section, such that the fluid for operating the small turbine flows therealong. 4. The system of claim 3, wherein the first circulation part is circulated by a single-phase fluid. 5. The system of claim 1, wherein the electric power produced during the normal operation of the nuclear power plant is supplied to an internal/external electric power system and an emergency power source. 6. The system of claim 5, wherein the electric power charged in the emergency power source is supplied as emergency power during the accident of the nuclear power plant. 7. The system of claim 1, wherein the electric power produced during the accident of the nuclear power plant is supplied as emergency power of the nuclear power plant. 8. The system of claim 6 or 7, wherein the emergency power is supplied as power for operating a safety system of the nuclear power plant during the accident of the nuclear power plant, opening and closing a valve for the operation of the safety system, monitoring the safety system, or operating the external reactor vessel cooling and electric power generation system. 9. The system of claim 1, wherein seismic design of seismic categories I, II or III is applied. 10. The system of claim 1, wherein safety classes 1, 2 or 3 are applied. 11. The system of claim 1, wherein the external reactor vessel cooling section is provided with a discharge pipe that connects the external reactor vessel cooling section and the power production section to each other such that the fluid of the external reactor vessel cooling section is applied to the power production section. 12. The system of claim 11, wherein the discharge pipe is provided with a first discharge portion through which at least part of the fluid excessively supplied to the power production section bypasses the small turbine and the small generator. 13. The system of claim 11, wherein the discharge pipe is further provided with a liquid gas separator that is connected to the discharge pipe such that only gas of the fluid is transferred to the power production section. 14. The system of claim 1, wherein the condensation heat exchange section is provided with a motor or pump that supplies a cooling fluid to the condensation heat exchange section to exchange heat with the fluid. 15. The system of claim 14, wherein the cooling fluid comprises air, pure water, seawater, or a mixture thereof. 16. The system of claim 1, wherein the condensed water storage section is disposed below the condensation heat exchange section to collect the condensed water generated in the condensation heat exchange section. 17. The system of claim 16, wherein the condensed water storage section is connected to the external reactor vessel cooling section through a pipe so that the condensed water is supplied to the external reactor vessel cooling section. 18. The system of claim 1, wherein the condensation heat exchange section or the condensed water storage section is provided with a vent portion through which non-condensable gas accumulated in the condensation heat exchange section or in the condensed water storage section is removed. 19. The system of claim 1, wherein at least part of a shape of the external reactor vessel cooling section includes a cylindrical shape, a hemispherical shape, and a double vessel shape, or a combination thereof. 20. The system of claim 1, wherein a pipe is connected to an in-containment refueling water storage tank (IRWST) such that refueling water is supplied to the external reactor vessel cooling section. 21. The system of claim 20, wherein the external reactor vessel cooling section is provided with a second discharge portion through which the refueling water supplied from the IRWST is discharged. 22. The system of claim 1, wherein a coating member is further provided to prevent corrosion of the reactor vessel. 23. The system of claim 22, wherein a surface of the coating member is chemically processed to increase a surface area. 24. The system of claim 1, further comprising a heat transfer member to smoothly transfer heat discharged from the reactor vessel. 25. The system of claim 24, wherein a surface of the heat transfer member is chemically processed to increase a surface area. 26. The system of claim 1, further comprising a core catcher provided inside the external reactor vessel cooling section to receive and cool corium when the reactor vessel is damaged. |
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summary | ||
description | The present application claims priority under 35 U.S.C. 119 to Korean Patent Application No. 10-2011-0067727, filed on Jul. 8, 2011, the disclosure of which is expressly incorporated by reference herein in its entirety. 1. Field of the Invention The present invention relates, in general, to an apparatus for measuring the thickness of an oxide layer formed on a cladding of each fuel rod of a nuclear fuel assembly using an eddy current sensor and, more particularly, to an apparatus for measuring the thickness of an oxide layer of a fuel rod, which includes driving means for providing movement in upward and downward, forward and backward, and leftward and rightward directions, a first probe for testing claddings of fuel rods placed in an outer side of a nuclear fuel assembly (hereinafter, “outer fuel rods”), and a second probe for testing claddings of fuel rods placed in an inner side of the nuclear fuel assembly (hereinafter, “inner fuel rods”). 2. Description of the Related Art During operation of a nuclear reactor, fuel rods disposed on each nuclear fuel assembly are immersed in a coolant/moderator in a reactor core. In light water reactors using zirconium or zircaloy cladding tubes for fuel rods, zirconium oxide (zirconia) is deposited on the fuel rods at a thickness of about 100 μm by a reaction between a coolant/moderator using water and zirconium in the cladding tubes. An adverse effect caused by the zirconia produced when heat is being transferred from the fuel rod cladding tube to the coolant/moderator and a metal loss which damages the structural integrity of the cladding tube, cause the thickness of the cladding tube to be reduced. As such, a restriction is placed on the maximum amount of oxide allowable to each fuel rod. Once the amount of oxide reaches this limit, the fuel rod must be replaced. Thus, the thickness measurement of the oxide layer is important for accurately evaluating the thermohydraulic performance of the fuel rod, estimating operational restrictions of the fuel rod, and estimating the longevity of the fuel rod. Generally, the thickness of the oxide layer of the fuel rod is measured by eddy current testing, and the damage to the fuel rod is checked by ultrasonic testing. Examples of the related art for measuring the thickness of the oxide layer of the fuel rod using eddy current testing include Korean Registered Utility Model No. 20-0339313, entitled “The probe fixture for eddy current testing of the RCCA of nuclear fuel type,” Korean Patent Application Publication No. 10-2004-0012065, entitled “Transfer apparatus of nuclear fuel rod for eddy current testing,” Korean Patent No. 10-0735213, entitled “Method for measuring oxide thickness underlying a ferromagnetic material on nuclear fuel rods,” and so forth. However, the eddy current testing disclosed in the related art requires much time to do a test because the nuclear fuel assembly should be disassembled and fuel rods disassembled from the nuclear fuel assembly should be tested one by one. As such, it cannot efficiently test the oxide layers of the fuel rods. Accordingly, the present invention has been made keeping in mind the problems of the related art, and an objective of the present invention is to provide a probe capable of continuously testing claddings of outer fuel rods of a nuclear fuel assembly without disassembling the nuclear fuel assembly while moving along a length of each fuel rod in upward and downward directions. Another objective of the present invention is to provide an apparatus for measuring the thickness of an oxide layer of a fuel rod, which includes first and second probes capable of testing claddings of outer and inner fuel rods of a nuclear fuel assembly without disassembling the nuclear fuel assembly. In order to achieve the above objectives, according to one aspect of the present invention, there is provided a probe, which comprises an eddy current sensor and a fuel rod transfer region so as to be able to continuously test a cladding of an outer fuel rod of a nuclear fuel assembly while moving in upward and downward directions. The probe can include a plurality of transfer regions having a semicircular lengthwise channel, and particularly, a fuel rod transfer region on which an eddy current sensor is mounted, a transfer roller disposed above the fuel rod transfer region, and transfer support regions disposed on either sides of the fuel rod transfer region. The probe can continuously test the cladding of the outer fuel rod of the nuclear fuel assembly while moving in upward and downward directions. According to another aspect of the present invention, there is provided an apparatus for measuring a thickness of an oxide layer of a fuel rod. The apparatus comprises a frame in which a cylinder driven in upward and downward directions is mounted, a first probe that is connected to one side of the cylinder in order to test claddings of outer fuel rods of a nuclear fuel assembly, and a second probe that is connected to another side of the cylinder in order to test claddings of inner fuel rods of the nuclear fuel assembly. The first probe can continuously test the claddings of outer fuel rods of the fixed nuclear fuel assembly using a first eddy current sensor 131 while moving in upward and downward directions. The second probe can include a strip and a second eddy current sensor, and test the claddings of inner fuel rods of the fixed nuclear fuel assembly using the second eddy current sensor while moving in forward and backward directions. According to the present invention as described above, the probe can continuously test the cladding of each outer fuel rod of the fixed nuclear fuel assembly while moving in upward and downward directions, so that it can examine the state of the oxide layer of the outer fuel rod all over rather than at a specific point. Further, the apparatus for measuring the thickness of an oxide layer of a fuel rod can simultaneously test the claddings of the outer and inner fuel rods of the nuclear fuel assembly without disassembling the nuclear fuel assembly, so that it can rapidly and efficiently measure the thickness of the oxide layer of each fuel rod. Reference will now be made in greater detail to an exemplary embodiment of the invention with reference to the accompanying drawings. Wherever possible, the same reference numerals will be used throughout the drawings and the description to refer to the same or like parts. In the following description, it is to be noted that, when the functions of conventional elements or the detailed description of elements related to the present invention may make the gist of the present invention unclear, a detailed description of those elements will be omitted. Nuclear fuel is arranged in a nuclear reactor on the basis of a nuclear fuel assembly. One nuclear fuel assembly is made up of tens or hundreds of fuel rods. One fuel rod is designed so that uranium pellets are covered with a zircaloy cladding tube having a thickness of 1 mm so as to be protected from external damage and to prevent radioactivity from leaking. The fuel rod has a diameter of about 9.5 mm, and an interval between the fuel rods is about 3.3 mm. As shown in FIGS. 1 to 6, a first probe 130 of the present invention includes a first eddy current sensor 131, a fuel rod transfer region 132, a transfer roller 133, transfer support regions 134, lateral support frames 135, and a first lower plate 136. Further, an apparatus for measuring the thickness of an oxide layer of a fuel rod is generally made up of a measurement unit 100, a transverse transfer unit 200 allowing the measurement unit 100 to move in leftward and rightward directions, a longitudinal transfer unit 300 allowing the measurement unit 100 to move in forward and backward directions, and a support unit 400 supporting the transverse transfer unit 200 and the longitudinal transfer unit 300. As shown in FIGS. 1A and 1B, the first probe 130 is a part that is configured to measure the thickness of an oxide layer of a fuel rod to be tested. The first eddy current sensor 131 is mounted on the first probe 130 so as to come into contact with a cladding of the fuel rod. The first eddy current sensor 131 detects an amount of eddy current induced on the fuel rod. Thereby, the thickness of the oxide layer of the fuel rod can be measured. In this embodiment, the eddy current sensor is a contact type of sensor, but it can be a contactless sensor or a proximity sensor. In detail, the fuel rod transfer region 132, in which a semi-circular channel is formed in a lengthwise direction at a predetermined length, is located between the transfer support regions 134, and the transfer roller 133 is mounted above the fuel rod transfer region 132 in a lengthwise direction. The transfer support regions 134, in each of which a semi-circular channel is formed in a lengthwise direction and is longer than that of the fuel rod transfer region 132, are located on either sides of the fuel rod transfer region 132, respectively. Further, the part configured to measure the thickness of the oxide layer of the fuel rod is fixed to the lateral support frames 135 connected to a lateral portion thereof, and the lateral support frames 135 are fixed by a first lower plate 136 connected to a bottom surface thereof. That is, the first lower plate 136 of the first probe 130 is connected to one side of a cylinder 120, as shown in FIG. 3. The fuel rod to be tested comes into contact with the fuel rod transfer region 132 on which the first eddy current sensor 131 is mounted. Here, the cylinder 120 is driven in upward and downward directions in parallel to the outer fuel rods of the nuclear fuel assembly, and thus a state of the oxide layer of each outer fuel rod is continuously measured by the first eddy current sensor 131 mounted on the fuel rod transfer region 132 of the first probe 130. The transfer roller 133 mounted above the fuel rod transfer region 132 in a lengthwise direction comes into contact with the fuel rod having an oxide layer to be tested, and the transfer support regions 134 located on the either sides of the fuel rod transfer region 132 come into contact with fuel rods located on left and right sides of the fuel rod having an oxide layer to be tested, respectively. Thus, the transfer roller 133 serves to guide the first probe 130 so that it does not deviate from its path while the first probe 130 connected to the cylinder 120 is driven in upward and downward directions to test the cladding of each outer fuel rod of the nuclear fuel assembly. The measurement unit 100 is a portion where probes for testing the claddings of fuel rods of nuclear fuel assembly are mounted. The measurement unit 100 includes a frame 110 in which the cylinder 120, which can be driven in upward and downward directions as shown in detail in FIG. 3, is mounted, the first probe 130 connected to one side of the cylinder 120 via the first lower plate 136, a second probe 140 connected to another side of the cylinder 120 via a second lower plate 143, and a transfer table 150 supporting the frame 110. As shown in FIG. 4, the second probe 140 includes a strip 141, a second eddy current sensor 142 mounted on one end of the strip 141, and the second lower plate 143 supporting the strip 141. In detail, the second lower plate 143 of the second probe 140 is connected to another side of the cylinder 120, as shown in FIG. 3. The measurement unit 100 is configured to be driven along first guide rails 210 of the transvers transfer unit 200 in leftward or rightward directions and along second guide rails 310 of the longitudinal transfer unit 300 in forward or backward directions. The apparatus 1 for measuring the thickness of an oxide layer of a fuel rod is positioned adjacent to the nuclear fuel assembly, and the strip 141, on which the second eddy current sensor 142 of the second probe 140 is mounted, is inserted between the fuel rods by the longitudinal transfer unit 300. At this time, the second eddy current sensor 142 detects an amount of eddy current induced on the fuel rod when coming into contact with the fuel rod, so that it can measure the state (e.g., thickness) of the oxide layer at a specific point of the fuel rod. In this embodiment, the second eddy current sensor is a contact type of sensor, but it can be a contactless sensor or a proximity sensor. Further, the strip 141 has a predetermined length, for example, a length of a side of the nuclear fuel assembly, and is configured to be able to be displaced in forward and backward directions by the longitudinal transfer unit 300. As such, when the measurement unit 100 is displaced in the forward direction by the longitudinal transfer unit 300, the strip 141 on which the second eddy current sensor is mounted is inserted into the nuclear fuel assembly, so that it is possible to examine the state of the oxide layer of the inner fuel rod of the nuclear fuel assembly. The transverse transfer unit 200 includes a first support bed 220 and the first guide rails 210 so that the measurement unit 100 including the first probe 130 and the second probe 140 can be driven in left and right directions. Sliding members 230 having a semicircular recess are mounted on a bottom surface of the first support bed 220, so that the transverse transfer unit 200 can be driven along the second guide rails 310 installed on the longitudinal transfer unit 300 in the forward and backward directions. In detail, the apparatus 1 for measuring the thickness of an oxide layer of a fuel rod is positioned adjacent to the nuclear fuel assembly. When the oxide layer of each outer fuel rod is to be measured by the first probe 130, the measurement unit 100 is moved by the transverse transfer unit 200 so that the fuel rod to be tested and the neighboring fuel rods located on left and right sides of the fuel rod to be tested are positioned in parallel to the fuel rod transfer region 132 and the transfer support regions 134. Further, when the oxide layer of each inner fuel rod is to be measured by the second probe 140, the second probe 140 is moved by the longitudinal transfer unit 300 so that the strip 141, on which the second eddy current sensor 142 is mounted, can be inserted between the fuel rods and contacted with an inner fuel rod to be measured. Since the longitudinal transfer unit 300 includes a second support bed 320 and the two second guide rails 310, the sliding members 230, each of which has a semicircular recess and is mounted on the bottom surface of the first support bed 220 of the transverse transfer unit 200, are allowed to be driven along the second guide rails 310 of the longitudinal transfer unit 300 in the forward and backward directions. In detail, when the inner fuel rods of the nuclear fuel assembly are tested by the second probe 140 mounted on the measurement unit 100, the strip 141 on which the second eddy current sensor 142 is mounted is positioned so as to be able to be inserted between the fuel rods by the transverse transfer unit 200, and then is inserted between the fuel rods by the longitudinal transfer unit 300. An amount of eddy current generated at the specific point where the inner fuel rod comes into contact with the second eddy current sensor 142 is detected so that the thickness of the oxide layer can be measured. The principle of measuring the thickness of the oxide layer on the basis of the eddy current sensor is as follows. When an alternating current flows to the eddy current sensor, an electromagnetic field is formed around the eddy current sensor. The electromagnetic field induces an eddy current on a surface of the fuel rod. If any defect is present on the surface of the fuel rod, the eddy current is distorted. When an amount of the distorted eddy current is measured, the type and size of the defect on the surface of the fuel rod can be detected. When measuring the thickness of the oxide layer, if the oxide layer has no defect, no eddy current is induced on the surface of the fuel rod. However, because the oxide layer is a nonconductor, the thicker the oxide layer becomes, the smaller the eddy current is induced on the surface of the fuel rod. This is called “lift-off.” The thickness of the oxide layer deposited on the cladding of the fuel rod can be measured by measuring this lift-off value. Although the exemplary embodiment of the present invention has been described 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. |
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summary | ||
abstract | A nuclear fusion system comprises a nuclear fusion device for providing heat energy, a capacitor for storing electrical energy for use by the nuclear fusion device in providing the heat energy, and an electrical conductor for carrying electrical energy from the capacitor to the nuclear fusion device, each of the nuclear fusion device, the capacitor and the conductor being located within a first chamber. The first chamber is located within a second chamber. A fluid is located between the first and second chambers, surrounds the nuclear fusion device, the capacitor and the conductor, and receives heat energy from each of the nuclear fusion device, the capacitor and the conductor, resulting in the fluid being heated. A thermal energy converter receives heated fluid from the second chamber. A super insulating material encloses the second chamber to reduce heat loss from the heated fluid to the cooler ambient. |
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description | This is a continuation, under 35 U.S.C. §120, of copending international application PCT/AT2005/000418, filed Oct. 21, 2005, which designated the United States; this application also claims the priority, under 35 U.S.C. §119, of Austrian application GM 780/2004, filed Oct. 27, 2004; the prior applications are herewith incorporated by reference in their entirety. Field of the Invention The invention relates to a cooling device component having a through-hole for carrying coolant, which comprises at least one heat shield made from tungsten, a tungsten alloy, a graphitic material or a carbidic material. First wall components for fusion reactors, such as for example diverters and limiters, which are exposed to very high loads of over 10 mW/m2, are a typical example of the use of cooling device components of this type. The region exposed to the plasma is referred to as a heat shield, while the component exposed to the plasma is referred to as the PFC (plasma facing component), and the material exposed to the plasma is referred to as the PFM (plasma facing material). PFMs must be plasma-compatible, must have a high resistance to physical and chemical sputtering, must have a high melting point/sublimation point and be as resistant as possible to thermal shocks. In addition, they must also have a high thermal conductivity, a low neutron activation and sufficient strength/fracture toughness, combined with good availability and acceptable costs. Tungsten, tungsten alloys (e.g. W-1 wt. % La2O3), graphitic materials (e.g. fiber-reinforced graphite) and carbidic materials (e.g. boron carbide) best satisfy this multi-faceted and in some cases contradictory profile of requirements. Since the energy flows act on these components for a prolonged period of time, cooling device components of this type are typically actively cooled. The dissipation of heat is assisted by heat sinks, for example made from copper or copper alloys, which are usually connected to the PFM. Cooling device components can have various designs. In this context, a distinction is drawn between plane tile, saddle and monobloc design. If a PFM tile with a planar connection surface is joined to the heat sink through which the coolant flows, this is referred to as a plane tile design. In the saddle design, a PFM body with a semicircular recess is joined to a tubular heat sink. The heat sink in this case has the function of producing the thermal contact between the heat-introduction side and the cooling medium and is exposed to cyclic, thermally induced loads resulting from the temperature gradient and the different expansion coefficients of the joining partners. In the monobloc design, a pipe carrying cooling water is surrounded by the PFM heat shield which has a closed through-hole. Whereas in the saddle and plane tile design, individual heat shield components can become detached from the heat sink on account of the cyclic, thermo-mechanical loading in use, the monobloc design precludes the loss of heat shield components for geometric reasons. However, the drawback of the monobloc design is that the PFM has to cope not just with thermally induced loads but also with additional mechanical loads. Additional mechanical loads of this nature can be produced by electromagnetically induced currents which flow in the components and interact with the surrounding magnetic field. This can give rise to high-frequency acceleration forces which have to be transmitted by the structures involved. In the plane tile and saddle design, these forces are transmitted via structural materials, but in the monobloc design they are transmitted by the PFM. However, tungsten, tungsten alloys, graphitic and carbidic materials have a low fracture toughness. An additional factor in the case of fiber-reinforced graphites is the relatively low strength. In addition, neutron embrittlement occurs in use, resulting in a further increase in the susceptibility of these materials to incipient cracking. Despite many years of expensive development work in the field of first wall components, the parts which are currently available do not optimally satisfy this profile of demands. This is one reason why large-scale industrial implementation of fusion technology remains far from imminent. Therefore, it is an object of the invention to provide a cooling device component of monobloc design (PFM heat shield with a through-hole) which suitably meets the demands resulting from both physical and mechanical stresses. This object is achieved by a cooling device component which comprises at least one heat shield made from tungsten, a tungsten alloy, a graphitic material or a carbidic material and is provided with a through-hole, and which component also comprises at least one structural part made from a material with a tensile strength at room temperature of >300 MPa and an electrical resistivity of >0.04 Ohm mm2m−1 as well at least one cooling pipe for carrying coolant. In accordance with the invention, the heat shield and the structural part are joined to one another by metallurgical joining or material bonding. The cooling device component according to the invention ideally satisfies the multi-faceted and in some cases contradictory profile of requirements presented in the description and therefore represents a simple solution to a long-standing problem. It has been found that if the strength of the structural part is >300 MPa and the electrical resistivity is >0.04 Ohm mm2m−1, it is possible to avoid both incipient cracking in the brittle PFM and loss of the heat shield. The heat shield preferably has a closed through-hole, for example in the form of a continuous bore. Preferred materials for the heat shield are CFC (carbon fiber reinforced graphite), pure tungsten and W-1 wt. % La2O3. Furthermore, it is expedient if for a projected length l and a projected width b of the heat shield the projected joining area between heat shield and structural part is >0.3·(l·b). The terms projected length, projected width and projected area in this context are to be understood as meaning perpendicular projection onto a planar surface. In the case of a line/area which is planar or rectilinear in form, the actual length, width and area correspond to the projected length, width and area. In the case of a curved line/area, the projected line/area is correspondingly reduced. To ensure component reliability for a prolonged period of time under stress for components which are subject to extremely high levels of loading, a projected joining area of >0.8 (l·b) is advantageous. Furthermore, it has proven expedient for the relative magnetic permeability of the structural part to be <1.2. Particularly suitable materials for the structural part in this context are Fe-, Ti- and Ni-based materials, which have the physical properties listed above. In this context, particular mention should be made of austenitic, ferritic, ferritic-martensitic steels and ODS (oxide dispersion strengthened) materials. To reduce the stresses in the material composite, it is advantageous for an interlayer, preferably with a thickness of 0.01 to 5 mm and made from a ductile material with a hardness of <300 HV, to be introduced between heat shield and structural part, in which case the interlayer particularly advantageously has a stress-relief notch. It has proven expedient for the interlayer to consist of copper or a copper alloy. Structuring of the joining area, preferably on the heat shield side, for example by means of a laser, also reduces the risk of cracking. Further expedient embodiments include a narrowing cross section of the structural part at increasing distance from the joining area, a planar design of the joining area and the introduction of a bevel or a rounded section in the edge region of the joining area/structural part. The joining between the heat shield and the structural part can be effected for example by soldering, by simultaneously and jointly backing both joining partners by casting or by backing the heat shield by casting with a thick, ductile layer, for example of copper or aluminum, followed by a subsequent joining process by soldering, electron beam welding or for example diffusion welding. Depending on the particular embodiment, the joining of the cooling pipe to the heat shield may take place before or after the process of joining to the structural part. This results in the following production variants, although the possible options are not restricted to the examples indicated. Variant 1 Back casting of the through-hole and a side face of the heat shield with a ductile material. Depending on the ductile material selected, this operation may take place in one process step or in two separate process steps. This is followed by the joining of the pipe to the component described above. The side face of the heat shield which has been provided with the ductile material is joined to the structural part, for example by means of soldering, EB welding or HIP (hot isostatic pressing). Variant 2 The through-hole and the surface of the heat shield on the side of the structural part are back cast with a ductile material. Depending on the ductile material selected, this can take place in one process step or in two separate process steps. This is followed by joining the component produced in this way to the structural part via the ductile layer (for example by means of soldering, EB welding or HIP). It is also possible for step 1 and step 2 to be combined in one process step. In this case, the structural part must have a higher melting point than the ductile material. Thereafter, the cooling pipe is joined to the component described above. The structural part in turn is mechanically connected or joined to a support element, for example made from steel. Joining can be effected for example by means of welding, while the mechanical connection can be effected by means of a rigid bolt connection by way of one or more fins. In the event of major differences in thermal expansion occurring in use, a sliding mechanical connection, which allows relative movement between structural part and support element, may also be advantageous. Other features which are considered as characteristic for the invention are set forth in the appended claims. Although the invention is illustrated and described herein as embodied in monobloc cooling device component, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims. The construction and method of operation of the invention, however, together with additional objects and advantages thereof will be best understood from the following description of specific embodiments when read in connection with the accompanying drawings. Referring now to the figures of the drawing in detail and first, particularly, to FIG. 1 thereof, the cooling device component comprises a heat shield 2, which is joined to a cooling pipe 4 via a ductile interlayer 7. The heat shield having the length l and the width b is in turn joined, at its joining area (6), to the structural part 3. The projected joining area amounts to l·b. The heat shield 2 and the structural part 3 are joined by a metallurgical joining process such as soldering, brazing, welding (e.g., electron beam welding) or pressure welding (e.g., hot isostatic pressing), where the materials of the parts that are being joined are physically or chemically bonded to each other or a molten phase occurs during the joining operation. FIG. 2 shows a cooling device component 1 in which heat shield 2 is joined to the structural part 3 by way of a ductile interlayer 13. In FIG. 3, the interlayer 13 has a stress-relief notch 8. FIG. 4 shows a cooling device component 1 in which the structural part 3 has a narrowing cross section. The edge regions of the joining area 6 are provided with radii 9 in FIG. 5 and with bevels in FIG. 6. Radii and bevels may be directed upwards or downwards. FIG. 7 shows a cooling device component 1 with a joined support element 11 by way of a weld joint 14. In FIG. 8, the support element 11 is mechanically connected to the structural part 3 by means of bolt connection 12 protruding through a fin 15. A cooling device component of monobloc design having a heat shield made from carbon fiber reinforced graphite (CFC) was produced as follows: Three CFC blocks with dimensions of 28 mm (ex-pitch direction) 25 mm (ex-PAN direction) and 20 mm (fiber direction) were provided with a through-hole in the form of a bore with a diameter of 14 mm. Then, the inside of the bore and one block side with dimensions of 25×20 mm were structured by laser means. In a subsequent step, OFHC copper was applied into the structured bore and/or onto the structured block surface by means of a casting process, with titanium being provided in the connecting area in order to ensure wetting. Thereafter, the OFHC copper on the block surface and in the bore was machined away to a thickness of between 0.5 and 1 mm. Thereafter, a pipe made from a copper-chromium-zirconium alloy with a diameter of approx. 12 mm was in each case introduced into the bore. A steel cuboid provided with a 5 μm thick nickel plating on all sides and with dimensions of 25×20×30 mm was in each case positioned on the block side provided with the OFHC copper layer. The assembly produced in this way was then placed in a steel can. The steel can was welded, evacuated and subjected to a HIP process at 550° C. and 1000 bar. After removal of the can, the cooling device component was tested by means of ultrasound. The connecting areas were free of defects. Thereafter, the cooling device component produced in this way was machined and slide-mounted on a steel structure by means of a bolt connection. A cooling device component of monobloc design with a heat shield made from tungsten was produced as follows: Three tungsten blocks with dimensions of 28×25×20 mm were in each case provided with a through-hole in the form of a bore with a diameter of 14 mm. The bore and a 25×20 mm side face of the tungsten block were back-cast with OFHC copper. Following the casting process, the OFHC copper both in the bore and on the block surface was machined away to a thickness of between 0.1 and 5 mm. The tungsten blocks produced in this way were then threaded onto a pipe made from a copper-chromium-zirconium alloy. A steel cuboid provided with a 5 μm thick nickel plating on all sides and with dimensions of 25×20×30 mm was in each case positioned on the block side provided with OFHC copper. Thereafter, this assembly was placed in a steel can. The steel can was welded, evacuated and subjected to a HIP process at 550° C. and 1000 bar. After removal of the can, an ultrasound test was carried out, and this did not reveal any defects. Thereafter, the cooling device component produced in this way was machined and fixed to a steel structure by means of electron beam welding. |
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description | The present invention relates to a nuclear power plant control system and a method of testing a nuclear power plant, and particularly relates to a nuclear power plant control system and a method of testing a nuclear power plant capable of conducting a test during an operation of the nuclear power plant. A nuclear power plant control system that controls a nuclear power plant is requested to regularly test each function to maintain high reliability. For example, a test of each function of the nuclear power plant control system that controls a nuclear power plant is conducted using test equipment during a periodical check when a nuclear reactor is shut down (for example, see Patent Literature 1). Recently, by applying a digital control device, software within a control device included in a nuclear power plant control system may maintain reliability through self-examination without a conventional periodical test. For this reason, a portion using a manual test is a hard-wired device (switch, cable, and the like) of an external unit or an input/output unit. Patent Literature 1: Japanese Patent Application Laid-open No. 2002-108443 However, since a length of a periodical check period is limited, it is preferable that a test of each function of a nuclear power plant control system be conducted during an operation of a plant as possible. In particular, in the US, a periodical check period is shortened, and a desire for permitting a test during an operation of a plant is strong. The invention is conceived in view of the above description, and an object of the invention is to provide a nuclear power plant control system and a method of testing a nuclear power plant capable of conducting a test during an operation of the nuclear power plant. According to an aspect of the present invention, a nuclear power plant control system includes: an operation unit which receives an operation for controlling a specific portion of a nuclear power plant; a notification unit which notifies that a control signal corresponding to the operation received by the operation unit arrives at a predetermined position on a path connected to the portion; and an inhabitation unit which inhibits the control signal from arriving at the portion in a position between the predetermined position on the path and the portion. In the nuclear power plant control system, a control signal may be inhibited from arriving at a control target, and a detection of a control signal at a predetermined position is notified even when the control signal is blocked. Accordingly, it is possible to test whether a control signal arrives at a predetermined position while an operation of a nuclear power plant is continued. Advantageously, in the nuclear power plant control system, the operation unit is provided in a central control room of the nuclear power plant, and the notification unit notifies an operator present in the central control room. In this embodiment, since an operator present in a central control room may conduct a test, it is possible to reduce personnel needed for the test. Advantageously, in the nuclear power plant control system, the inhabitation unit inhibits the control signal from arriving at the portion when a signal indicating a state of being on test is received. In this embodiment, by transmitting a signal indicating a state of being on test, it is possible to conduct a test while an operation of a nuclear power plant is continued. Advantageously, in the nuclear power plant control system, the inhabitation unit is a majority circuit that causes a control signal to arrive at the portion only when a plurality of control signals are received through different paths. In this embodiment, by not transmitting signals from a plurality of systems, it is possible to conduct a test while an operation of a nuclear power plant is continued. Advantageously, in the nuclear power plant control system, the inhabitation unit is a majority circuit that causes a nuclear reactor trip signal to arrive at the portion when receiving the nuclear reactor trip signal as a control signal from detecting units of at least N (N is an integer greater than or equal to 2) systems among detecting units of a plurality of systems, the operation unit is provided to correspond to each of the detecting units of the plurality of systems, and an OR operation is performed on a control signal indicating that the operation unit is operated and a signal output by the detecting unit of a system corresponding to the operation unit before the control signal is detected at the predetermined position. In this embodiment, by transmitting a signal from a system, it is possible to conduct a test associated with a nuclear reactor trip while an operation of a nuclear power plant is continued. According to another aspect of the present invention, a method of testing a nuclear power plant includes: receiving an operation through an operation unit which receives an operation for controlling a predetermined portion of a nuclear power plant; detecting a control signal corresponding to the operation received by the operation unit at a predetermined position on a path connected to the portion; notifying that the control signal arrives at the predetermined position; and inhibiting the control signal from arriving at the portion in a position between the predetermined position on the path and the portion. In the method of testing a nuclear power plant, a control signal may be inhibited from arriving at a control target, and a detection of a control signal at a predetermined position is notified even when the control signal is blocked. Accordingly, it is possible to test whether a control signal arrives at a predetermined position while an operation of a nuclear power plant is continued. A nuclear power plant control system and a method of testing a nuclear power plant according to the embodiment have an effect of being able to conduct a test during an operation of a nuclear power plant. Hereinafter, embodiments of a nuclear power plant control system and a method of testing a nuclear power plant according to the invention will be described in detail based on drawings. It should be noted that the invention is not limited to the embodiments. In addition, elements in the embodiments include elements that may be easily assumed by those skilled in the art, substantially identical elements, and so-called equivalents. First Embodiment In addition, a configuration of a nuclear power plant control system according to a first embodiment will be described with reference to FIG. 1. FIG. 1 is a diagram illustrating a schematic configuration of a nuclear power plant control system 1 according to the first embodiment. As illustrated in FIG. 1, the nuclear power plant control system 1 includes an operation terminal 10, a control panel 20, a control device 30, and a control target device 40. The operation terminal 10 and the control panel 20 are installed inside a central control room 6. The operation terminal 10 receives various operations for operating a nuclear power plant, and shows various types of information associated with an operation condition of the nuclear power plant. The control panel 20 includes a control button 21, and transmits a control signal for controlling the control target device 40 to the control device 30 in response to an operation received by the control button 21. The control panel 20 may be integrated with the operation terminal 10. In addition, the control button 21 may be a physical button, and may be a virtual button displayed on a screen. The control target device 40 provides a predetermined function based on a control signal transmitted by the control panel 20. For example, the control target device 40 corresponds to a valve, a pump, a heater, and the like provided in the nuclear power plant. The control device 30 performs predetermined arithmetic processing based on an input signal, and outputs a signal corresponding to a result of the operation. In addition, the control device 30 has a function of conducting a test with respect to an operation of the control button 21 without suspending an operation of the nuclear power plant. As a configuration associated with the function, the control device 30 includes a digital input unit (DI) 31, a digital output unit (DO) 32, and a control signal inhibition unit 33. The DI 31 receives a control signal transmitted from the control panel 20. The DO 32 transmits a control signal to the control target device 40. The control signal inhibition unit 33 includes a NOT circuit 33a and a AND circuit 33b, and inhibits a control signal received by the DI 31 from being transmitted from the DO 32 to the control target device 40 based on an operation performed on the operation terminal 10. Specifically, the operation terminal 10 is provided with a test permission button 11, and the operation terminal 10 transmits, as a test permission signal, “1” (High) or “0” (Low) to the control signal inhibition unit 33 in response to an operation performed on the test permission button 11. Herein, it is presumed that a test permission signal “1” indicates a state of being on test with respect to an operation of the control button 21, and a test permission signal “0” indicates a state of being in normal operation. The test permission button 11 may be a physical button, and may be a virtual button displayed on a screen. The control signal inhibition unit 33 inverts a test permission signal in the NOT circuit 33a, and inputs the inverted test permission signal to the AND circuit 33b. In addition, the control signal inhibition unit 33 inputs a control signal received by the DI 31 to the AND circuit 33b. The AND circuit 33b performs an AND operation on the inverted test permission signal and the control signal, and outputs a result of the operation to the DO 32. Thus, when the test permission signal is “1”, that is, during a test with respect to an operation of the control button 21, the control signal inhibition unit 33 inhibits the control signal received by the DI 31 from being transmitted from the DO 32 to the control target device 40. In addition, when the test permission signal is “0”, that is, during a normal operation, the control signal inhibition unit 33 transmits the control signal received by the DI 31 from the DO 32 to the control target device 40. In addition, the control device 30 is configured such that a control signal received by the DI 31 branches before being input to the control signal inhibition unit 33, and the branching control signal is transmitted to the operation terminal 10. The operation terminal 10 includes a notification lamp 12, and turns on the notification lamp 12 when a control signal is transmitted from the control device 30. That is, the nuclear power plant control system 1 is configured such that when a control signal transmitted in response to an operation of the control button 21 is output from the DI 31 to the control signal inhibition unit 33, the corresponding information is notified by the notification lamp 12 regardless of whether the control signal inhibition unit 33 blocks the control signal. In the nuclear power plant control system 1 having a configuration described above, when a test target portion 7 (a portion at a front of and at a rear of the DI 31) illustrated in FIG. 1 is needed to be tested with respect to an operation of the control button 21 during an operation of the nuclear power plant, an operator inside the central control room 6 performs an operation as follows. First, the operator operates the test permission button 11 so that a signal indicating a state of being on test with respect to an operation of the control button 21 is transmitted from the operation terminal 10 to the control device 30. Next, the operator operates the control button 21, and verifies a notification being performed by the notification lamp 12 in response to the operation. In this instance, a control signal is blocked by the control signal inhibition unit 33, and does not arrive at the control target device 40. Then, after verifying a notification of the notification lamp 12, the control button 21 is returned to an original state, and the test permission button 11 is returned to an original state. Through the operation described in the foregoing, even when the nuclear power plant is in operation, it is possible to test whether the test target portion 7 is normally functioning without changing a control state of the control target device 40. In addition, the nuclear power plant control system 1 has a merit of being able to reduce equipment and personnel needed for a test. A test of a nuclear power plant control system has been conducted using test equipment attached to a control device included in the nuclear power plant control system. For this reason, in a conventional nuclear power plant control system, an operator who operates test equipment is needed to be disposed near a control device in addition to an operator inside a central control room to conduct a test during an operation of a nuclear power plant. In the nuclear power plant control system 1, since a test may be conducted while an operator is inside the central control room 6, an operator may not be disposed near the control device 30, and test equipment is not needed. Next, an operation of the control device 30 illustrated in FIG. 1 will be described with reference to FIG. 2. FIG. 2 is a flowchart illustrating an operation of the control device 30. As illustrated in FIG. 2, in response to a control signal being received by the DI 31 of the control device 30 as step S10, the control device 30 turns on the notification lamp 12 of the operation terminal 10 as step S11. Then, when a test permission signal transmitted from the operation terminal 10 is “1”, that is, during a test with respect to an operation of the control button 21 (Yes in step S12), the control signal inhibition unit 33 inhibits a control signal from being transmitted to the control target device 40 as step S13. On the other hand, when a test permission signal transmitted from the operation terminal 10 is “0”, that is, during a normal operation (No in step S12), the control signal inhibition unit 33 transmits a control signal to the control target device 40 as step S14. As described in the foregoing, in the first embodiment, the control signal inhibition unit 33 is provided to block a control signal based on an operation of the test permission button 11, and the notification lamp 12 is provided to notify a state of a control signal irrespective of an operation of the control signal inhibition unit 33. Accordingly, it is possible to conduct a test of the nuclear power plant control system 1 while an operator inside the central control room 6 continues to operate the nuclear power plant. Second Embodiment In the first embodiment, an example of inhibiting a control signal from arriving at the control target device 40 during a test by transmitting a signal indicating a state of being on test from the operation terminal 10 to the control signal inhibition unit 33 by operating the operation terminal 10 has been described. However, a control signal may be inhibited from arriving at a device to be controlled during a test by another mechanism. Thus, in a second embodiment, an example of inhibiting a control signal from arriving at a device to be controlled during a test by using a majority circuit will be described. First, a configuration of a nuclear power plant control system according to the second embodiment will be described with reference to FIG. 3. FIG. 3 is a diagram illustrating a schematic configuration of a nuclear power plant control system 2 according to the second embodiment. As illustrated in FIG. 3, the nuclear power plant control system 2 includes operation terminals 50a to 50d, detectors 60a to 60d, control devices 70a to 70d, a trip control device 80, and a control rod driving mechanism 90. The operation terminals 50a to 50d are installed inside a central control room 6. The detectors 60a to 60d include a sensor used to detect an occurrence of an event using a trip of a nuclear reactor, and a threshold calculation unit used to threshold-calculate a detected value of the sensor, respectively. Then, when a detected value of the sensor exceeds a threshold, the detectors 60a to 60d transmit a nuclear reactor trip signal, which is a control signal used to execute a trip of the nuclear reactor, to a control device 70. As described in the foregoing, the nuclear power plant control system 2 is configured such that four systems detect an occurrence of an event using a trip of the nuclear reactor independently of one another. The trip control device 80 opens a power-supply circuit to the control rod driving mechanism 90 in response to a detection state in the detectors 60a to 60d. Specifically, the trip control device 80 includes a majority circuit 81, and the majority circuit 81 opens the power-supply circuit to the control rod driving mechanism 90 when a nuclear reactor trip signal is transmitted from at least two systems among the four systems. As described above, by providing the majority circuit 81, it is possible to properly operate the nuclear power plant control system 2 even when a portion of the systems has an error. In addition, as described below, the majority circuit 81 functions as a control signal inhibition unit that inhibits the power-supply circuit from being opened to the control rod driving mechanism 90 during a test. When the power-supply circuit is opened in the trip control device 80 and power is supplied, the control rod driving mechanism 90 trips the nuclear reactor by driving a control rod. The control devices 70a to 70d perform predetermined arithmetic processing based on an input signal, and output a signal corresponding to a result of an operation. As a configuration associated with the nuclear reactor trip signal, the control device 70a includes a DI 71a, an OR circuit 72a and a DO 73a. As a configuration associated with the nuclear reactor trip signal, the control device 70b includes a DI 71b, an OR circuit 72b, and a DO 73b. As a configuration associated with the nuclear reactor trip signal, the control device 70c includes a DI 71c, an OR circuit 72c, and a DO 73c. As a configuration associated with the nuclear reactor trip signal, the control device 70d includes a DI 71d, an OR circuit 72d, and a DO 73d. The DIs 71a to 71d receive a nuclear reactor trip signal transmitted from the detectors 60a to 60d, respectively. The OR circuits 72a to 72d output a signal, obtained by performing an OR operation on a nuclear reactor trip signal received by the DIs 71a to 71d and a pseudo nuclear reactor trip signal transmitted from the operation terminals 50a to 50d, to the DOs 73a to 73d, respectively. The DOs 73a to 73d transmit, as a nuclear reactor trip signal, a signal output from the OR circuits 72a to 72d to the trip control device 80, respectively. That is, from the control devices 70a to 70d, both a nuclear reactor trip signal transmitted from the detectors 60a to 60d and a pseudo nuclear reactor trip signal transmitted from the operation terminals 50a to 50d are transmitted as a nuclear reactor trip signal to the trip control device 80. The operation terminals 50a to 50d receive various operations for operating a nuclear power plant, and show various types of information associated with an operation condition of the nuclear power plant. As a configuration associated with a test of the nuclear power plant control system 2, the operation terminal 50a includes a pseudo trip button 51a and a notification lamp 52a. As a configuration associated with a test of the nuclear power plant control system 2, the operation terminal 50b includes a pseudo trip button 51b and a notification lamp 52b. As a configuration associated with a test of the nuclear power plant control system 2, the operation terminal 50c includes a pseudo trip button 51c and a notification lamp 52c. As a configuration associated with a test of the nuclear power plant control system 2, the operation terminal 50d includes a pseudo trip button 51d and a notification lamp 52d. The operation terminal 50a corresponds to the same system as that of the control device 70a. The operation terminal 50b corresponds to the same system as that of the control device 70b. The operation terminal 50c corresponds to the same system as that of the control device 70c. The operation terminal 50d corresponds to the same system as that of the control device 70d. It should be noted that the pseudo trip buttons 51a to 51d may be physical buttons, and may be virtual buttons displayed on a screen. In response to the pseudo trip buttons 51a to 51d being operated, the operation terminals 50a to 50d transmit a pseudo nuclear reactor trip signal to a corresponding system according to a performed operation. The pseudo nuclear reactor trip signal transmitted from an operation terminal 50 arrives at the trip control device 80 through the control device 70 of the same system. However, as described above, the trip control device 80 opens the power-supply circuit to the control rod driving mechanism 90 when signals are transmitted from at least two systems. For this reason, even when one of the pseudo trip buttons 51a to 51d is operated and a pseudo nuclear reactor trip signal is transmitted, the power-supply circuit is not opened to the control rod driving mechanism 90 by the trip control device 80. In addition, the nuclear power plant control system 2 is configured such that a signal transmitted from the DOs 73a to 73d branches before arriving at the trip control device 80, and the branching signal is transmitted to the operation terminal 50 of the same system. In response to receiving the branching signal, the operation terminals 50a to 50d turn on a notification lamp 52 of the respective operation terminals 50a to 50d. That is, the nuclear power plant control system 2 is configured such that when a signal transmitted in response to an operation of the pseudo trip buttons 51a to 51d is output from the DOs 73a to 73d to the trip control device 80, the corresponding information is notified by the notification lamps 52a to 52d even when the trip control device 80 is blocking a nuclear reactor trip signal. In the nuclear power plant control system 2 having a configuration described above, when a test target portion 8 (a portion where a signal is output from the DOs 73a to 73d) illustrated in FIG. 3 is needed to be tested during an operation of the nuclear power plant, an operator inside the central control room 6 performs an operation as follows. First, an operator operates one of the pseudo trip buttons 51a to 51d to cause a pseudo nuclear reactor trip signal to be transmitted from the operation terminal 50, and verifies that a notification lamp, among the notification lamps 52a to 52d, corresponding to an operated button is turned on. The operator verifies lighting by performing a similar operation for another button of the pseudo trip buttons 51a to 51d. Through the operation described above, even when the nuclear power plant is in operation, it is possible to test whether the test target portion 8 is normally functioning for each system without opening the power-supply circuit to the control rod driving mechanism 90. In addition, in the nuclear power plant control system 2, an operator may conduct a test while being inside the central control room 6, and thus test equipment and an operator who operates the test equipment are not needed similarly to the nuclear power plant control system 1. Next, an operation of the nuclear power plant control system 2 illustrated in FIG. 3 will be described with reference to FIG. 4. FIG. 4 is a flowchart illustrating an operation of the nuclear power plant control system 2. As illustrated in FIG. 4, in response to receiving a signal (nuclear reactor trip signal or pseudo nuclear reactor trip signal) as step S20, the control device 70 transmits the received signal to the trip control device 80 as step S21. In this instance, the nuclear power plant control system 2 turns on a notification lamp, among the notification lamps 52a to 52d, of a system corresponding to the received signal as step S22. Then, when signals are not received from two systems or more (No in step S23), the trip control device 80 inhibits a nuclear reactor trip signal from being transmitted to the control rod driving mechanism 90 as step S24. On the other hand, when signals are received from two systems or more (Yes in step S23), the trip control device 80 transmits a nuclear reactor trip signal to the control rod driving mechanism 90 as step S25. As described in the foregoing, in the second embodiment, a pseudo nuclear reactor trip signal may be transmitted to the trip control device 80 including the majority circuit 81 for each system, and the notification lamps 52a to 52d are provided to notify a circumstance of transmitting a signal to the trip control device 80 irrespective of an operation of the majority circuit 81. Accordingly, it is possible to conduct a test of the nuclear power plant control system 2 while an operator inside the central control room 6 continues to operate the nuclear power plant. It should be noted that a configuration of the nuclear power plant control system described in each embodiment above may be arbitrarily changed within a scope not departing the spirit of the invention. For example, configurations of the nuclear power plant control systems described in the respective embodiments above may be arbitrarily combined together. In addition, in each embodiment above, a notification lamp is turned on to notify a test result. However, as a scheme of notifying a test result, various schemes such as a scheme using a notification sound, a scheme of displaying a result using a character or a symbol on a screen may be used. 1, 2 NUCLEAR POWER PLANT CONTROL SYSTEM 6 CENTRAL CONTROL ROOM 7, 8 TEST TARGET PORTION 10, 50a to 50d OPERATION TERMINAL 11 TEST PERMISSION BUTTON 12, 52a to 52d NOTIFICATION LAMP 20 CONTROL PANEL 21 CONTROL BUTTON 30, 70a to 70d CONTROL DEVICE 31, 71a to 71d DI 32, 73a to 73d DO 33 CONTROL SIGNAL INHIBITION UNIT 33a NOT CIRCUIT 33b AND CIRCUIT 40 CONTROL TARGET DEVICE 51a to 51d PSEUDO TRIP BUTTON 60a to 60d DETECTOR 72a to 72d OR CIRCUIT 80 TRIP CONTROL DEVICE 81 MAJORITY CIRCUIT 90 CONTROL ROD DRIVING MECHANISM |
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abstract | A method for imaging an object is presented. The method includes acquiring a first projection image of the object using a source and a detector, positioning a scatter rejecting aperture plate between the object and the detector, and acquiring a second projection image of the object with the scatter rejecting aperture plate disposed between the object and the detector. A scatter image of the object is generated based on the first projection image and the second projection image, and stored for subsequent imaging. A volumetric CT system for imaging an object is also presented. The system is configured to acquire a plurality of projection images of the object from a plurality of plurality of projection angles. |
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043483569 | claims | 1. In a nuclear reactor vessel support system including a nuclear reactor vessel having an outwardly extending support ring, a nuclear reactor containment structure surrounding the vessel, the containment structure having an inwardly extending support ledge aligned with and attached to the support ring whereby the vessel weight is transmitted to the support ledge, the improvement comprising a box ring interposed in load bearing relationship directly between the support ring and support ledge to transmit the vessel weight from the support ring to the support ledge, said box ring having rigid cylindrical side walls and flat rigid ring-shaped upper and lower walls to enclose and define an annular space directly between the support ledge and the support ring to limit heat flow therebetween. 2. The improvement of claim 1 wherein the top and bottom of the box ring are low alloy steel and the side walls thereof are of relatively thin Inconel. 3. The improvement of claim 1 including radial keys in the top and bottom of the box ring and mating keyways in the bottom of the reactor vessel support ring and in the top plate of the reactor vessel support ledge. 4. The improvement of claim 1 wherein holddown bolts are provided that extend through said support ring and support ledge. 5. The improvement of claim 1 including a coolant header defined by a seal collar spaced from the outside edge of the support ring and a coolant channel extending from the coolant header to the reactor containment cavity between the reactor vessel and the support ledge. 6. The improvement of claim 5 wherein said seal collar comprises a ring of metal plate having an angular cross-section with a horizontal and a vertical flange, said horizontal flange outer edge in sealing engagement with said support ring and said vertical flange outer edge in sealing engagement with said reactor containment. 7. The improvement of claim 5 wherein a plurality of radial keyways commuting with said coolant header are provided in the top surface of said support ledge for receiving radial keys also in keyway engagement with said box rings, said keys including channels in alignment with said keyway for passing coolant flow. |
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048797367 | abstract | An x-ray examination apparatus has an x-ray image intensifier and an x-ray film changer mounted so that the x-ray film changer can be moved to an exposure position, in front of the x-ray image intensifier, for producing x-ray pictures of an examination subject. The x-ray image intensifier has a holder on which the x-ray film changer is movably mounted so that, when in the exposure position, the film changer has the same image axis as that of the x-ray image intensifier. The film changer is pivotable out of the exposure position by approximately 90.degree. to a standby position. To avoid having the x-ray film changer in the standby position prevent access to a patient on an examination table, and to permit the x-ray image intensifier to be adjusted to any arbitrary position without being impeded by the film changer in the standby position, the examination apparatus has a holder for the film changer in the form of a bearing which at least partially surrounds the x-ray image intensifier. The bearing is disposed in a plane which is substantially perpendicular to the image access, and the film changer is displaceable along the bearing. |
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description | The present application is a continuation of U.S. patent application Ser. No. 14/630,141, filed on Feb. 24, 2015, now issued as U.S. Pat. No. 9,551,048, which is a continuation of U.S. patent application Ser. No. 13/705,012, filed on Dec. 4, 2012, now issued as U.S. Pat. No. 8,961,647, which is a continuation of U.S. patent application Ser. No. 12/312,089, filed on Sep. 16, 2009, now issued as U.S. Pat. No. 8,323,373, which in turn is a U.S. national stage application under 35 U.S.C. § 371 of PCT Application No. PCT/US2007/071233, filed Jun. 14, 2007, which claims priority to U.S. Provisional Patent Application No. 60/854,725, filed Oct. 27, 2006, the entireties of which are incorporated herein by reference. The present invention relates generally to the art of aluminum alloys. More specifically, the invention is directed to the use of powder metallurgy technology to form aluminum composite alloys which maintain their high performance characteristics even at elevated temperatures. The invention accomplishes this through the use of nanotechnology applied to particulate materials incorporated within the aluminum alloy. The resulting alloy composite has high temperature stability and a unique linear property/temperature profile. The alloy's high temperature mechanical properties are achieved by a uniform distribution of nano-sized alumina particulate in a superfine grained, nano-scaled aluminum matrix which is formed via the use of superfine atomized aluminum powder or aluminum alloy powder as raw material for the production route. The matrix can be pure aluminum or one or more of numerous aluminum alloys disclosed hereinbelow. Conventional aluminum materials exhibit many desirable properties at ambient temperatures such as, light weight and corrosion resistance. Moreover, they can be tailor-made for various applications with relative ease. Thus aluminum alloys have dominated the aircraft, missile, marine, transportation, packaging, and other industries. Despite the well known advantages of conventional aluminum alloys, their physical properties can be degraded at high temperatures, for example above 250° C. Loss of strength is particularly noticeable, and this loss of strength is a major reason why aluminum alloys are generally absent in demanding high temperature applications. In place of aluminum, the art has been forced to rely on much more expensive alloys such as those containing titanium or tungsten as the main alloying metal. Various attempts have been made to overcome the deficiencies of aluminum alloys at high temperatures. For example, U.S. Pat. No. 5,053,085 relates to “High strength, heat resistant aluminum based alloys” having at least one element from an M group consisting of V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Mo, W, Ca, Li, Mg and Si and one element from X group consisting of Y, La Ce, Sm, Nd, Hf, Ta, and Mm (Misch metal) blended to various atomic percentage ratios. These various alloy combinations produce an amorphous, microcrystalline phase, or microcrystalline composite dispersions through rapid solidification of molten aluminum. Rapid solidification of the aluminum is accomplished through melt spinning techniques which produce ribbon or wire feed stock. The ribbon or wire feed stock can be crushed and consolidated into billets for fabrication into various products through conventional extrusion, forging, or rolling technologies. Mechanical alloying is another attempt to produce high strength aluminum alloys. Nano particle strengthening of metal matrix materials is achieved in high-energy ball mills by reducing the particulates to fine dispersoids which strengthen the base alloy. A major problem associated with this technology is the uneven working of the particulates. A given volume of material is grossly over or under processed which leads to flaws in the final structure. U.S. Pat. No. 5,688,303 relates to a mechanical alloying process which incorporates the use of rolling mill technology to allegedly improve the homogenization of the mechanical alloying. Some of the largest obstacles to mechanical alloying technology include lack of ductility and powder handling issues. Handling of the mechanically alloyed powders is dangerous since the protective oxide is removed from the aluminum powder which then becomes pyrophoric. Aluminum powder without the protective oxide will ignite instantaneously when exposed to atmosphere so extreme caution is required during the handling of the powder blend. Moreover, the use of high energy ball mills is very expensive and time consuming which results in higher material processing costs. Other attempts to improve high temperature physical properties include the incorporation of additives. U.S. Pat. No. 6,287,714 relates to “Grain growth inhibitor for nanostructured materials”. Boron nitride (BN) is added as a grain growth inhibitor for nanostructure materials. This BN addition is added as an inorganic polymer at about 1% by weight and is uniformly dispersed at the grain boundaries which are decomposed during the heat treat temperature of the nanostructure material. U.S. Pat. No. 6,398,843 relates to “Dispersion-strengthened aluminum alloy” for dispersion strengthened ceramic particle aluminum or aluminum alloys. This patent is based on blending ceramic particles (alumina, silicon carbide, titanium oxide, aluminum carbide, zirconium oxide, silicon nitride, or silicon dioxide) with a particle size<100 nm. U.S. Pat. No. 6,630,008 relates to “Nanocrystalline metal matrix composites, and production methods” which involves using a chemical vapor deposition (CVD) process to fluidize aluminum powder which is coated with aluminum oxide, silicon carbide, or boron carbide then hot consolidated in the solid-state condition using heated sand as a pressure transmitting media. U.S. Pat. No. 6,726,741 relates to aluminum composite material and manufacture based on an aluminum powder and a neutron absorber material, and a third particle. Mechanical alloying is used in the manufacturing process. U.S. Pat. No. 6,852,275 relates to a process for production of inter-metallic compound-based composite materials. The technology is based on producing a metal powder preform and pressure infiltrating aluminum which results in a spontaneous combustion reaction to form inter-metallic compounds. Rapid solidification processing (RSP) technology is another method employed to produce fine metallic powders. However, RSP has high costs associated with atomization of the high soluble alloying elements, powder production rates, chemistry control, and recovery steps needed in order to maintain the amorphous and nano size microstructures. The other major obstacle with RSP is the difficulty in fabrication of the materials. These processes, while promising, have heretofore failed to address the long felt needs of manufacturing high temperature aluminum alloys on a commercial scale. Thus traditional, non-aluminum based alloys continue to dominate the high temperature alloy markets. The present invention overcomes the deficiencies of the prior art by taking advantage of the oxide coating which naturally forms during the atomization process to manufacture aluminum powder and by taking advantage of processing of powders with a particle size distribution below 30 μm. It is known that oxides exist on atomized aluminum powder regardless of the type of atomization gas used to manufacture. See, “Metals Handbook Ninth Edition Volume 7—Powder Metallurgy” by Alcoa Labs (FIG. 1). An indication of the oxide content can be estimated by measuring the oxygen content of the aluminum powder. Generally the oxygen content does not significantly change whether air nitrogen, or argon gases are used to manufacture the powder. As aluminum powder surface area increases (aluminum powder size decreases) the oxygen content increases dramatically, indicating a greater oxide content. The average thickness of the oxide coating on the aluminum powders is an average of about 5 nm regardless of the type of atomization gas but is independent of alloy composition and particle size. The oxide is primarily alumina (Al2O3) with other unstable compounds such as Al (OH) and AlOOH. This alumina oxide content is primarily controlled by the specific surface area of the powder. Particle size and particle morphology are the two main parameters which influence the specific surface area of the powder (>the surface area) respectively the more irregular (>the surface area) the higher the oxide content. With conventional aluminum powder sizes having a Particle Size Distribution (PSD) of <400 μm the particle shape/morphology becomes a very important factor towards controlling the oxide content since the irregular particle shape results in a greater surface area thus a higher oxide content. With a particle size <30 μm the effect of particle morphology has less influence on oxide content since the particles are more spherical or even ideal spherical in nature. Generally, the oxide content for various atomized aluminum particle sizes varies between about 0.01% up to about 4.5% of alumina oxide. The present invention targets starting aluminum or aluminum alloy powders with particles of <30 μm in size which will provide between 0.1-4.5 w/o alumina oxide content. The invention provides for hot working the desired PSD aluminum or aluminum alloy powder which produces in situ transversal nano-scaled grain size in the range of about 200 nm (a grain size reduction of factor 10.times.). Secondly the hot work operation produces in situ evenly distributed nanoscaled alumina oxide particles (the former oxide skins of the particles) with a thickness of max. 3-7 nm, resulting in high superior strength/high temperature material compared to conventional aluminum ingot metallurgy material. The superior mechanical properties are a result of the tremendous reduction in grain size and the uniform distribution of the nano-scale alumina oxide in the ultra fine grained aluminum matrix. It is accordingly an aspect of the invention to use this 0.1-4.5 w/o nano particle alumina reinforced aluminum composite material as a structural material for higher strength and higher temperature in a variety of market applications. This nano size aluminum/alumina composite structure shall be produced without the use of mechanical alloying but only by the use of a aluminum or aluminum alloy powder with a particle size distribution <30 μm m resulting in a nano-scaled microstructure after hot working. It is another aspect of this invention to obtain additional strength by the addition of a ceramic particulate material to the nano aluminum composite matrix material to obtain even greater strength, higher modulus of elasticity (stiffness), lower coefficient of thermal expansion (CTE), improved wear resistance, and other important physical properties. This ceramic particulate addition may include inter alia ceramic compounds such as alumina, silicon carbide, boron carbide, titanium oxide, titanium dioxide, titanium boride, titanium diboride, silicon oxide, silicon dioxide, and other industrial refractory compositions. It is another aspect of the invention to add boron carbide particulate to this nano aluminum composite matrix for neutron absorption for the storage of spent nuclear fuel as set forth in U.S. Pat. No. 5,965,829 entitled “Radiation Absorbing Refractory Composition” issued Oct. 12, 1999 (the '829 patent) which is hereby incorporated by reference in its entirety. It is another aspect of the invention to include other aluminum alloys such as high solubility elemental compositions in order to have a dual strengthened material through precipitation of fine intermetallic compounds through rapid solidification (in situ) of super saturated alloying element melt along with the nano-scale alumina particles uniformly dispersed throughout the microstructure after the hot work operation to produce the final product. It is another aspect of this invention to have technology based on a bimodal particle size distribution which will exhibit uniform micro structural control without the use of mechanical alloying technology. Control of microstructure size and homogeneity dictates the high performance of the composite material. It is another aspect of the invention to tailor the mechanical and physical properties for various market applications by changing the alloy composition of the nano aluminum/alumina composite matrix, the type of ceramic particulate addition, and the amount of ceramic particulate addition to the nano aluminum/alumina metal matrix composite material. These aspects and others set forth below, are achieved by a process for manufacturing a nano aluminum/alumina metal matrix composite characterized by the steps of providing, an aluminum powder having a natural oxide formation layer and an aluminum oxide content between about 0.1 and about 4.5 wt. % and a specific surface area of from about 0.3 and about 5.0 m.sup.2/g, hot working the aluminum powder, and forming thereby a superfine grained matrix aluminum alloy, and simultaneously forming in situ a substantially uniform distribution of nano particles of alumina throughout said alloy by redistributing said aluminum oxide, wherein said alloy has a substantially linear property/temperature profile. The aspects of the invention are also achieved by an ultra-fine aluminum powder characterized by from about 0.1 to about 4.5 wt. % oxide content with a specific surface area of from about 0.3 to about 5.0 m2/g, which is hot worked at a temperature ranging from about 100° C. to about 525° C. depending on the recrystallization temperature of a particular aluminum alloy composition to refine grain size and homogenize the nano particle reinforcement phase of the metal matrix composite system. In carrying out the invention, the first step is selection of aluminum powder size. The present invention focuses on the particle size distribution (PSD) of the atomized aluminum powder which is not used for conventional powder metal technology. In fact the trend in aluminum P/M industry is to use coarser fractions of the PSD—typical in the d50 size of 50 .mu.m-400 μm range because of atomization productivity, recovery, lower cost, superior die fill or uniform pack density and the desire to have low oxide powder. Most commercial applications seek to reduce the oxide content especially in the press and sinter near-net-shape aluminum P/M parts for automotive and other high volume applications. Manufacturers of powder and end-users want the lower oxide aluminum powder since it is extremely difficult to perform liquid phase sintering and obtain a metallurgical particle to particle bond which is necessary to obtain theoretical densities and high mechanical properties with acceptable ductility values with oxide on the powder grain boundaries. The prior grain boundary oxide network results in low fracture toughness, low strength, and marginal ductility. Efforts have been made to reduce the alumina oxide but this oxide coating on the aluminum powder is extremely stable in all environments and is not soluble in any solvent. This fact leads the press and sinter near-net-shape industry and the high performance aerospace industry aluminum PM industry to purchase low oxide powder material. In total contrast to the above noted industry criteria, the present invention employs superfine aluminum powder PSD (by industrial definition a PSD <30 μm) which results in alumina oxide content in the 0.1-4.5 w/o range, which is the oppose side of the spectrum. The invention includes taking the superfine powder and hot working the material below the recrystallization temperature of the alloy which further reduces the transverse grain size by a factor of 10 to a typical grain size of e.g. about 200 nm. The effect of the starting powder particle size is illustrated in FIG. 2 which shows the effect of 1 μm, 10 μm, and <400 μm aluminum powder extruded at 350° C. The hot work operation evenly distributes nanoscale alumina oxide particles (the former 3-7 nm oxide skin of the aluminum powder) uniformly throughout the microstructure as illustrated in FIG. 3 and circled in the micrograph. This ultra fine grain size and the nanoscale alumina particles combination results in a dual strengthening mechanism. The nanoscale alumina oxide particles pin the grain boundaries and inhibit grain growth to maintain the elevated mechanical property improvement of the composite matrix material. In certain embodiments, the oxide is redistributed into uniformly dispersed nano alumina particles intermixed with inter-metallic compounds. In certain embodiments, the inter-metallic compounds have a particle size of from about 2 to about 3 μm. It has been found that increasing the alumina oxide content of one specific type of powder by 50% does not result in higher mechanical properties compared to the original powder. Increasing the oxide content by 100% or more may result in problems during consolidation process. During powder treatment to increase the alumina oxide content only the thickness of the oxide layer can be increased which results in bigger dispersoids in the matrix after hot working. To increase the strengthening mechanism of grain boundary pinning, which is the designated positioning of nano-scaled dispersoids (alumina particles, the former oxide layer of the starting powder) along the grain boundaries of the microstructure, it is desirable to bring more fine particles into the structure. This can be realized by using a finer starting powder, or a powder with a higher specific surface area. By considering the particle size distribution together with the specific surface area of the starting powders, the mechanical properties of the hot worked material can be predicted. Powders with a higher specific surface will generally result in better mechanical properties compared to powders with a lower specific surface area. As can be seen in FIG. 4 powder sample #9 has roughly the same specific surface area as powder sample #5, although the PSD of sample #9 is much coarser than the PSD of sample #5. The mechanical properties correlate with the specific surface area, not with the PSD of the powders (FIG. 5). This figure shows UTS vs particle size distribution and specific surface area (test results of mechanical properties obtained on test specimen containing 9% of boron carbide particulate). Mechanical properties (UTS) correlate with BET not with the d50. Different powders with specific surface areas in the range between 0.3-5.0 m2/g were hot worked by extrusion at 400° C. into rods with a diameter of 6 mm which had been used for the production of tests specimen for tensile tests. The results are shown in the table and chart of FIGS. 6(a) and 6(b), respectively. This demonstrates that the finer the particle distribution (the higher the surface area) the better the mechanical properties. Powders were produced via gas atomization using confined nozzle systems and classified to required PSD via air classification. Afterwards, compacts were produced, by extrusion@400° C., R 11:1. High temperature tensile tests were made after 30 min. soak time@testing temperature. An example of the aluminum particle size used for the development is illustrated in FIG. 7. This graph illustrates PSD and as can be seen, the d50 is about 1.27 μm with d90 about 2.27 μm, which is extremely fine. Attached is a Scanning Electron Microscope (SEM) photograph (FIG. 8) “Picture of ultra fine atomized Al powder D50-1.2 μm” and Transmission Electron Micrograph (TEM). See FIG. 9, “Picture of ultra fine atomized Al powder D50-1.3 .mu.m” which illustrates the spherical shape of the powder. As shown therein, the hum marker (SEM) respectively the 0.2 μm marker (TEM) is a reference to verify the particle size of the powder. Since the aluminum powder in the particle size range is considered spherical it is easier to mathematically model and predict the oxide content. When modeling the oxide thickness and comparing the actual value of the oxide by dissolving the matrix alloy, there is good correlation that documents the targeted aluminum oxide content of the invention. Another characteristic of the powder is the very high surface area of the resulting PSD and the oxygen content as an indicator of the total oxide content of the starting raw material. The purchase specification to assure superior performance shall include the alloy chemistry, particle size distribution, surface area, and oxygen content requirements. FIG. 10 illustrates the unique linear property/temperature profile of the high temperature nano composite aluminum alloy of the invention. The figure shows UTS (Rm) vs. temperature, 1.27 μm (d50) powder grade, consolidated via direct extrusion@350° C., R=11:1, 30 min. soak time at testing temperature before testing. The typical processing route to manufacture the material for this invention is to fill the elastomeric bag with the preferred particle size aluminum powder, place the elastomeric top closure in the mold bag, evacuate the elastomeric mold assembly to remove a air and seal the air tube, cold isostatic press (CIP) using between 25-60,000 psi pressure, dwell for 45 seconds minimum time at pressure, and depressurize the CIP unit back to atmospheric pressure. The elastomeric mold assembly is then removed from the “green” consolidated billet. The billet can be vacuum sintered to remove both the free water and chemically bonded water/moisture which is associated with the oxide surfaces on the atomized aluminum powder. Care must be taken not to overheat the billet or approach the liquid phase sintering temperature in order to prevent grain growth and obtain optimum mechanical properties. The last operation is to hot work the billet to obtain full density, achieve particle to particle bond, and most importantly disperse the nano alumina particles uniformly throughout the microstructure. A preferred hot work method is to use conventional extrusion technology to obtain the full density, uniformly dispersed nano particle aluminum/alumina oxide composite microstructure. Direct forging or direct powder compact rolling technology could also be used as a method to remove the oxide from the powder and uniformly disperse the alumina oxide throughout the aluminum metal matrix. It is highly preferred to keep the extrusion temperature below the re-crystallization temperature of the alloy in order to obtain the optimum structure and optimum mechanical properties. FIGS. 11(a) and 11(b) are SEM photo micrographs which illustrate the importance of the extrusion temperature in order to increase the flow stress to mechanically work the material to obtain the desired microstructure. In photo micrograph FIG. 11(a) are visible the uniformly dispersed nano-alumina oxide particles in the newly formed grains. The nano particle alumina oxide particles are visible even inside the grain and at the grain boundaries which typically is done through the mechanical alloying process methods. The second photo micrograph FIG. 11(b) shows the larger grain size and the structure does not exhibit the same degree of work or the nano particles inside the grains. To further demonstrate the significance of extrusion temperature in obtaining the desired microstructure for optimum mechanical properties, outlined below are typical mechanical properties of the nano aluminum/alumina composite material at various extrusion temperatures on tensile data at room temperature and 350° C. test temperatures. MechanicalVarious Billet Extrusion TemperaturesProperties350° C.400° C.450° C.500° C.Room TemperatureUTS - Mpa/310 (44.95)305 (44.25)290 (42.05)280 (40.60)(KSI)Yield - Mpa/247 (35.82)238 (34.51)227 (32.91)213 (30.88)(psi)Elongation % 9.0%10.0%10.0%10.9%1100 Alumi-124 (18.00)N/AN/AN/Anum/UTS350° C. Test TemperatureUTS-Mpa/186 (26.97)160 (23.20)169 (24.50)160 (23.20)(KSI)Yieid-Mpa/156 (22.62)145 (21.00)150 (21.75)150 (21.75)(KSI)Elongation10.7%10.4% 9.5%10.0% These are excellent mechanical properties for a 4.5% nano alumina particle reinforced 1100 series superfine grained aluminum material compared with conventional ingot metallurgy 1100 series aluminum technology. Further, these results demonstrate the advantages of the superfine grained microstructure in combination with the small amount of nano particle aluminum/alumina materials compared to various conventional alloys and the concept of adding other ceramic particulate or rapid solidification of super saturated alloy elements in the aluminum matrix. As mentioned above, one of aspects of this invention is to add a ceramic particulate to the nano aluminum/alumina composite matrix. One of the driving forces to the development of this new technology was the need for a high temperature matrix material to add boon carbide particle to expand the field of application of U.S. Pat. No. 5,965,829. It was a goal to develop a high temperature aluminum boron carbide metal matrix composition material suitable to receive structural credit from the US Nuclear Regulatory Commission for use as a basket design for dry storage of spent nuclear fuel applications. With elevated temperature mechanical properties of the aluminum boron carbide composite, designers can take advantage of the light weight/high thermal heat capacity of aluminum metal matrix composites compared to the industry standard stainless steel basket designs. In Europe, designers typically use boronated stainless steel but the areal density is low, the upper limit for the B10 isotope being 1.6% content, alloy density is high, and the thermal conductivity and thermal heat capacity is low compared to aluminum based composites. The aluminum-based composites of the present invention do not suffer from these shortcomings. Another driving force behind the development of an aluminum boron carbide metal matrix higher temperature composite, in addition to the market need for such a material, was the experience with extruding up to 33 wt % boron carbide composite materials in a production environment, including the techniques described in U.S. Pat. No. 6,042,779 entitled “Extrusion Fabrication Process for Discontinuous Carbide Particulate Metal Matrix. Composites and Super Hypereutectic Al/Si Alloys,” issued on Mar. 28, 2000 (the '779 patent) and which is hereby incorporated by reference in its entirety. This extrusion technology could allow designers the freedom of design to extrude to net-shape a variety of hollow tube profiles in order to maximize packing density, add flux traps, and lower manufacturing cost. A particular use for the addition of ceramic particulate to the nano particle aluminum/alumina high temperature matrix alloy is the addition of nuclear grade boron carbide particulate. All of the tramp elements for the alloy matrix material such as Fe, Zn, Co, Ni, Cr, etc. are held to the same tight restrictions and the boron carbide particulate is readily available in accordance to ASTM C750 as outlined in the above described U.S. Pat. No. 5,965,829. The boron carbide particulate particle size distribution is similar to that outlined in the '829 patent. An exception to the teaching of the '829 patent is the use of high purity aluminum powder with the new particle size distribution as described above. The typical manufacturing route for the composite of the invention is first blending the aluminum powder and boron carbide particulate materials, followed by consolidation into billets using CIP plus vacuum sinter technology as outlined in the above referenced patent. In a preferred embodiment, the extrusion is carried out in accordance with the teaching of U.S. Pat. No. 6,042,779 (the '779 patent), which is hereby incorporated by reference in its entirety. Since this is an elevated temperature aluminum metal matrix composite material it was found necessary to change the temperature of the extrusion die, container temperature, and billet temperature in order to maintain the desired properties. In general it is desirable that the die face pressure be increased by about 25% over previously employed standard metal matrix composite materials. In order to overcome the higher flow stress of the nano particle aluminum/alumina composite matrix alloy, the extrusion press must be sized about 25% larger in order to extrude the material. Extrusion die technology is capable of these higher extrusion pressures without experiencing failure of collapse of the extrusion die. An example of the new high temperature nano particle aluminum/alumina plus boron carbide at a 9% boron carbide reinforcement level and the resulting typical mechanical properties and physical properties are outlined below. Property25° C.100° C.200° C.300° C.350° C.Description(70° F.)(212° F.)(392° F.)(572° F.)(662° F.)UTS-MPa/238/34.5208/30.2166/24.4126/18.3116/16KSIYield -194/28.1164/23.8150.21.7126/18.2105/15Mpa/KSIElongation11%10%9.0%8.0%8.0%%Modulus of 83/12.2 81/11.9 73/10.763/9.2 55/7.9ElasticityMPa/MPSIThermal184185184183Conductivity(W/m-K)Thermal106107106107Conductivity(BTU/ft-hr-° F.)Specific0.9931.0531.0991.121HeatJ/g-° C.Specific0.2370.2520.2690.280Heat (BTU/lb-° F.)Notes:Tensile coupons were machined and tested in accordance in ASTM E8 &ASTM E 21Thermal conductivity tested in accordance to ASTM E 1225Specific heat tested in accordance to ASTM E 1461 |
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051868873 | abstract | There is disclosed an inspection apparatus for peripheral surfaces of nuclear fuel pellets which includes a handling unit, an image pick-up device, a judging device and a sorting unit. The handling unit holds a prescribed number of nuclear fuel pellets in a line and rotates the same on their axes. The image pick-up device is disposed adjacent to the handling unit, and picks up image data as to the peripheral surfaces of the nuclear fuel pellets. The judging device is operably connected to the image pick-up device, and analyzes the image data outputted from the image pick-up device to output judging signals. The sorting unit is operably connected to the judging device and separates defective pellets from non-defective pellets based on the judging signals. The sorting unit includes a plurality of sorting members and actuators. The sorting members are disposed adjacent to the handling unit so as to correspond to the nuclear fuel pellets, respectively. The actuators are operably connected to the judging device and the sorting members, and operate the sorting members based on the judging signals. |
041397789 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT In nuclear reactor systems it is necessary to store fuel assemblies before and after the fuel assemblies are used in the nuclear reactor. The invention described herein provides apparatus capable of storing such fuel assemblies without the fuel assembly developing unacceptable stresses therein. Referring to FIG. 1, a typical ductless fuel assembly 10 comprises fuel elements 12. Fuel elements 12 may be hollow cylindrical metallic tubes filled with nuclear fuel as is well known in the art. Fuel elements 12 are held at their upper end by upper end support 14 and at their lower end by lower end support 16. Upper end support 14 and lower end support 16 are metal supports that are capable of maintaining proper alignment of fuel elements 12 within fuel assembly 10 and of providing a mechanism for attachment during transportation and reactor operation. Both upper end support 14 and lower end support 16 have openings 18 therein for accommodating alignment pins. During storage, fuel assembly 10 is supported from its upper end by clamp 20 which is a metal support conforming to the shape of upper end support 14. Once fuel assembly 10 has been placed in clamp 20 by typical fuel handling apparatus, clamp 20 may be made secure by tightening the winged nut and bolt arrangement 22 or other suitable fastening means. Clamp 20 is attached at the end opposite nut and bolt arrangement 22 to center post 24 by means well known in the art such as by sleeve 26 which may be welded to center post 24. Center post 24 may be a hollow cylindrical metal column supported vertically by its attachment to base 28 along with supporting metal struts 30. Base 28 may be either a metal plate fixed to the floor of the storage facility or it may be the floor itself. Center post 24 has four symmetrically disposed clamps 20 arranged in conjunction with struts 30 such that at least four fuel assemblies 10 may be supported from one center post 24. While fuel assembly 10 is supported from its upper end support 14 by clamp 20, the lower end support 16 rests on swivel base 32. Referring now to FIG. 2, swivel base 32 comprises first plate 34, second plate 36, third plate 38, ball 40, circular pins 42, and diamond pins 44. Third plate 38 may be a rectangular carbon steel plate approximately one inch thick and fastened to base 28 by appropriate means such as first screws 46. Second plate 36 is a rectangular carbon steel plate approximately one inch thick with its corners removed to expose the corners of third plate 38. Second plate 36 has a first hole 48 drilled in the center thereof that is complimentary to a similar center drilled second hole 50 in third plate 38 for accommodating a hard steel ball 40. Ball 40 which may be approximately 1.5 inches in diameter rests in second hole 50 of third plate 38 while second plate 36 rests on ball 40 such that ball 40 fits into first hole 48. When second plate 36 rests on ball 40 in this arrangement, second plate 36 is separated from third plate 38 by a gap 52 which may be approximately 0.125 inch wide. A foam rubber or urethane sheet 54 with a hole therein corresponding to first hole 48 and second hole 50 may be inserted into gap 52 between second plate 36 and third plate 38 so as to prevent foreign material from becoming lodged between the plates. Second plate 36 is, thereby, capable of moving relative to third plate 38 by pivoting or rotating on ball 40. A first plate 34 which may be manufactured of nylon or stainless steel and may be approximately 0.25 inch thick is attached to the top of second plate 36 by means well known to those skilled in the art such as second screws 56. First plate 34 provides a mechanism for isolating contact of lower end support 16 from second plate 36 to thereby avoid corrosion of fuel assembly 10 because contact of fuel assembly 10 with the carbon steel of second plate 36 may cause corrosion of fuel assembly 10. Hardened round pins 42 and diamond pins 44 are press fitted through first plate 34 and into second plate 36. Round pin 42 and diamond pin 44 are capable of fitting into openings 18 in lower end support 16 so as to engage lower end support 16 and support the weight of fuel assembly 10. Round pin 42 has a diameter 58 which may be approximately 0.8 inch and diamond pin 44 has a length 60 across farthest tips which is equal to diameter 58 while line 62 denotes the distance across the short tips of diamond pin 44 which is less than diameter 58. When fuel assembly 10 is placed on swivel base 32, a round pin 42 engages an opening 18 while a diamond pin 44 located across the center of swivel base 32 from the round pin 42 engages a similar opening 18 thereby preventing rotation of fuel assembly 10 with respect to plates 34 and 36 of swivel base 32. Line 62 being shorter than length 60 allows a fuel assembly 10 to be positioned on a round pin 42-diamond pin 44 set even though the corresponding openings 18 are separated by a distance slightly different from the distance between the pins while still preventing rotation of fuel assembly 10. A first round pin 42-diamond pin 44 set is located on first plate 34 to accommodate a fuel assembly 10 having a 14 .times. 14 array of fuel elements while a second set of pins are located at 45.degree. from the first set so as to accommodate a fuel assembly 10 having a 15 .times. 15 array of fuel elements 12 thereby providing swivel base 32 with the capability of supporting either type of fuel assembly. Furthermore, first plate 34 and second plate 36 have a centered hole 64 therethrough that allows clamp 20 to be aligned with the center of swivel base 32 by plumb bob or other such methods. When a typical fuel assembly 10 has been placed into clamp 20, lower end support 16 rests on swivel base 32 with a diamond pin 44 and corresponding round pin 42 engaging openings 18 of lower end support 16. While in this position the fuel assembly 10 may not be completely vertically aligned because of misalignment in clamp 20 or for other common reasons such as thermal bending. Since a typical fuel assembly 10 may be approximately 156 inches in length and weigh approximately 1500 pounds, this slight misalignment may cause severe stresses in the fuel assembly. However, because the fuel assembly 10 rests on swivel base 32, first plate 34 and second plate 36 together move relative to third plate 38 by revolving on ball 40 thereby allowing fuel assembly 10 to become aligned in a nonstressed position. When so moving, second plate 36 may either compress sheet 54 or allow sheet 54 to expand slightly. While in the prior art, a slight misalignment may have produced a high stress level in a fuel assembly without a swivel base thereby approaching the situation of a long slender column with two clamped ends, the present invention approaches the condition of a column with one end clamped and the other end pivoted. Therefore, the invention provides a fuel assembly storage rack having a swivel base with plates capable of relative motion for supporting fuel assemblies while minimizing the stresses in the fuel assembly. While there is described what is now considered to be the preferred embodiment of the invention, it is, of course, understood that various other modifications and variations will occur to those skilled in the art. The claims, therefore, are intended to include all such modifications and variations which fall within the true spirit and scope of the present invention. For example, ball 40 may be replaced with a hemispherical member that would rest with its flat side on third plate 38 while its hemispherical side would extend into first hole 48. |
040595390 | summary | The present invention relates to an improved form of UN. More particularly, it is concerned with a single-phase uranium-zirconium mononitride composition which is characterized by an enhanced thermal stability at temperatures of 1600.degree. C. and above in comparison to uranium mononitride. Uranium mononitride (UN) is an attractive nuclear fuel because of its high uranium density (14.32), high melting point (2850.degree. C. at 2.5 atmospheres), and usefully high thermal conductivity (0.06 at 1000.degree. C.). Because of its high uranium content, UN can, for example, be substituted for UO.sub.2 and occupy approximately 30 percent less volume at an equivalent uranium content. This combination of properties makes UN an excellent candidate fuel for use in high-temperature (fuel temperatures greater than 1500.degree. up to 1700.degree. C.) fast reactors, especially those designed to operate in outer space. A limiting factor which mitigates against its use in such and similar context is its dissociation into liquid uranium and nitrogen gas under reduced pressures at these higher temperatures. It is accordingly a principal object of this invention to provide a modified uranium mononitride composition which does not dissociate at a temperature in the range 1500.degree.-1700.degree. C. In order to derive maximum benefit from UN it is generally held desirable to make it available for use as a homogeneous single-phase structure as opposed to a multiphase structure in which disruptive phase changes may occur during use or where homogeneity is difficult to obtain. It is therefore an additional object of this invention to provide a homogeneous single-phase uranium mononitride composition having improved thermal stability. A further object is to provide a modified sintered UN compact having a lowered dissociation constant or rate of decomposition at 1600.degree. C. to 1700.degree. C. in comparison to pure UN. |
claims | 1. A source collimator for use with a line source in single photon emission computed tomography (SPECT), which source collimator comprises:first and second side walls and first and second end walls and septa positioned between said first and second side walls and parallel to said end walls, wherein said walls are comprised of corrugated segments of hexagonal holes and wherein the collimation angle of the source collimator is between about 10° and about 14°. 2. The source collimator of claim 1, wherein the collimation angle of the source collimator is between about 11° and about 13°. 3. The source collimator of claim 2, wherein the collimation angle of the source collimator is about 12°. 4. The source collimator of claim 1, wherein the corrugated segments of hexagonal holes further include lead side walls. 5. A line source holder for use in single photon emission computed tomography (SPECT), which line source holder comprises:a holder with first and second interconnecting slots, said first slot adapted to hold a base and said second slot adapted to hold a source collimator and a radiation absorber;a base adapted to fit into said first slot of said holder and having a groove adapted to hold a line source of radiation;a line source of radiation;a source collimator; anda radiation absorber, said source collimator and said radiation absorber adapted to fit into said second slot of said holder; said source collimator comprising:first and second side walls and first and second end walls and septa positioned between said first and second side walls and parallel to said end walls, wherein said walls are comprised of corrugated segments of hexagonal holes and wherein the collimation angle of the source collimator is between about 10° and about 14°. 6. The line source holder of claim 5, wherein the collimation angle of the source collimator is between about 11° and about 13°. 7. The line source holder of claim 6, wherein the collimation angle of the source collimator is about 12°. 8. The line source holder of claim 5, wherein the corrugated segments of hexagonal holes further include lead side walls. 9. A SPECT system, comprising:a gantry;a scintillation detector mounted to said gantry; anda transmission line source assembly including a plurality of radiation line sources mounted in corresponding line source grooves in an assembly base, each radiation line source comprising:a holder with first and second interconnecting slots, said first slot adapted to hold a base and said second slot adapted to hold a source collimator and a radiation absorber, a holder base adapted to fit into said first slot of said holder and having a groove adapted to hold a line source of radiation, a source collimator and a radiation absorber, said source collimator and said radiation absorber adapted to fit into said second slot of said holder, said source collimator comprising:first and second side walls and first and second end walls and septa positioned between said first and second side walls and parallel to said end walls, wherein said walls are comprised of corrugated segments containing holes therein. 10. A SPECT system as set forth in claim 9, wherein the collimation angle of the source collimator is between about 10° and about 14°. 11. A SPECT system as set forth in claim 10, wherein the collimation angle of the source collimator is between about 11° and about 12°. 12. A SPECT system as set forth in claim 11, wherein the collimation angle of the source collimator is about 12°. |
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claims | 1. An installation for treating articles with radiation, the installation comprising a structure having pivotally mounted thereon an inlet starwheel and an outlet starwheel respectively arranged facing an inlet and an outlet of a shielded enclosure in which there are mounted at least one pivotal treatment starwheel and at least one electron emitter in the vicinity of the treatment starwheel, the starwheels being provided with article-gripper means, the shielded enclosure having internal shielded partitions arranged as baffles, wherein the installation includes at least one linear conveyor extending inside the shielded enclosure facing the inlet or the outlet, the linear conveyor having a transporter extending as an elongate ring around a straight shielded wall extending in a longitudinal direction of a straight portion of the transporter in order to form a baffle, the straight shielded wall facing said inlet or outlet and having a length greater than a width of said inlet or outlet. 2. An installation according to claim 1, including a linear inlet conveyor and a linear outlet conveyor extending inside the shielded enclosure respectively facing the inlet and the outlet, the linear conveyors each comprising a transporter surrounding a shielded wall extending in a longitudinal direction of the transporter in order to form a baffle at the inlet and a baffle at the outlet, respectively. 3. An installation according to claim 1, wherein at least one intermediate starwheel is mounted inside the shielded enclosure between the treatment starwheel and the linear conveyor. 4. An installation according to claim 1, wherein the shielded internal partitions separate the treatment starwheel from the linear conveyor while leaving an opening for passing the articles. |
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051805499 | claims | 1. A double enclosure top nozzle subassembly for a nuclear fuel assembly, said top nozzle subassembly comprising: (a) an upper structure including a top plate and an outer sidewall enclosure rigidly connected to and depending from an outer peripheral edge of said top plate; (b) a lower structure including a lower adaptor plate and an inner sidewall enclosure rigidly connected to and upstanding from an outer peripheral edge of said lower plate, said lower adapter plate being disposed below said top plate, said inner sidewall enclosure being disposed within said outer sidewall enclosure, said inner and outer sidewall enclosures being movable in sliding contacting relationship relative to one another so as to permit movement of said top plate toward and away from said lower adapter plate; (c) interengaging means on respective upper and lower peripheral edges of said inner and outer sidewall enclosures for defining stops which limit the movement of said top plate and lower adaptor plate away from each other so as to retain said outer and inner sidewall enclosures in said sliding contacting relationship with one another and prevent movement of said lower peripheral edge of said outer sidewall enclosure below said lower adaptor plate, said interengaging means on said respective upper and lower peripheral edges of said inner and outer sidewall enclosures also defining sliding surfaces between said inner and outer sidewall enclosures being maintained in continuous contact with one another as said top plate and lower adapter plate move toward and away from one another; and (d) a plurality of resiliently-yieldable biasing devices disposed within said inner and outer sidewall enclosures and extending between and engaging said top plate and said lower adapter plate, said devices being movable between compressed and expanded states in response respectively to application and removal of a hold-down force on said upper structure in the direction of said lower structure for respectively permitting and causing movement of said inner and outer sidewall enclosures relative to one another and said top plate toward and away from said lower adapter plate and thereby said top nozzle subassembly between compressed and expanded conditions. an inwardly-projecting flange on said lower edge of the outer sidewall enclosure defining a contact surface engaged with an exterior surface of said inner sidewall enclosure; and an outwardly-projecting flange on said upper edge of said inner sidewall enclosure defining a contact surface engaged with an interior surface of said outer sidewall enclosure, said flanges overlapping with one another to define said stops. said lower adapter plate has a depression formed in a topside of said lower adapter plate along each of said sides and approximately midway between said corners thereof; said top plate has a pair of guide grooves defined in the underside of said top plate along each side and adjacent said corners thereof; and each of said leaf springs has a generally U-shaped configuration composed of a lower bight portion and upper end portions connected to and extending upwardly from said lower bight portion, said leaf spring being seated at said lower bight portion within one of said depressions of said lower adapter plate and at said opposite upper end portions within said guide grooves of said top plate. (a) an upper structure including a top plate and an outer sidewall enclosure rigidly connected to and depending from an outer peripheral edge of the top plate; (b) a lower structure including a lower adapter plate and an inner sidewall enclosure rigidly connected to and upstanding from an outer peripheral edge of said lower adapter plate, said lower adapter plate being disposed below said top plate, said inner sidewall enclosure being disposed below said top plate, said sidewall enclosure, said inner and outer sidewall enclosures being movable in sliding contacting relationship relative to one another so as to permit movement said top plate toward and away from said lower adapter plate; (c) interengaging means on respective upper and lower peripheral edges of said inner and outer sidewall enclosures for defining stops which limit the movement of said top plate and lower adapter plate away from each other so as to retain said outer and inner sidewall enclosures in said sliding contacting relationship with one another and prevent movement of said lower peripheral edge of said outer sidewall enclosure to below said lower adapter plate, said interengaging means on said respective upper and lower peripheral edges of said inner and outer sidewall enclosures also defining sliding surfaces between said inner and outer sidewall enclosures being maintained in continuous contact with one another as said top plate and lower adapter plate move toward and away from one another; and (d) a plurality of resiliently-yieldable biasing devices disposed within said inner and outer sidewall enclosures and extending between and engaging said top plate and the lower adapter plate, said devices being movable between compressed and expanded states in response respectively to application and removal of a hold-down force on said upper structure in the direction of said lower structure for respectively permitting and causing movement of said inner and outer sidewall enclosures relative to one another so as to move said top plate toward and away from said lower adapter plate and thereby said top nozzle subassembly between compressed and expanded conditions. an inwardly-projecting flange on said lower edge of the outer sidewall enclosure defining a contact surface engaged with an exterior surface of said inner sidewall enclosure; and an outwardly-projecting flange is an inwardly-projecting flange on said upper edge of said inner sidewall enclosure defining a contact surface engaged with an interior surface of said outer sidewall enclosure, said flanges overlapping with one another to define said stops. said lower adapter plate has a depression formed in a topside of said lower adapter plate along each of said sides and approximately midway between said corners thereof; said top plate has a pair of guide grooves defined in the underside of said top plate along each side and adjacent said corners thereof; and each of said leaf springs has a generally U-shaped configuration composed of a lower bight portion and upper end portions connected to and extending upwardly from said lower bight portion, said leaf spring being seated at said lower bight portion within one of said depressions of said lower adapter plate and at said opposite upper end portions within said guide grooves of said top plate. 2. The top nozzle subassembly as recited in claim 1, wherein said interengaging means includes: 3. The top nozzle subassembly as recited in claim 1, wherein said top plate of said upper structure includes an annular body. 4. The top nozzle subassembly as recited in claim 3, where said outer peripheral edge of said top plate is defined by a rim projecting outwardly from an upper peripheral edge of said annular body so as to define an annular cavity surrounding said annular body for receiving an upper peripheral edge of said outer sidewall enclosure. 5. The top nozzle subassembly as recited in claim 3, wherein said top plate includes a plurality of indentations defined along the periphery and underside of said annular body in spaced relation to one another, said indentations facing outwardly of said body and downwardly therefrom, the portions of said annular body between said indentations forming downwardly protruding tabs along the periphery of said body for attaching said top plate to an upper peripheral edge of said outer sidewall enclosure. 6. The top nozzle subassembly as recited in claim 5, wherein said top plate further includes a plurality of pins inserted through said upper peripheral edge of said outer sidewall enclosure and said tabs of said annular body for securing said outer sidewall enclosure and said annular body together. 7. The top nozzle subassembly as recited in claim 5, wherein said inner sidewall enclosure of said lower structure has a plurality of upper edge portions spaced apart by notches defined between said upper edge portions, said upper edge portions and said notches being capable of mating respectively with said spaced indentations and tabs in said annular body of said top plate of said upper structure when said top nozzle subassembly is in said compressed condition. 8. The top nozzle subassembly as recited in claim 1, wherein said top plate of said upper structure and said lower adapter plate of said lower structure are generally rectangular in configuration having four sides defining four corners. 9. The top nozzle subassembly as recited in claim 8, wherein said biasing devices are a plurality of leaf springs disposed between said top plate and said lower adapter plate and movable between expanded and compressed states. 10. The top nozzle subassembly as recited in claim 9, wherein: 11. The top nozzle subassembly as recited in claim 1, wherein each of said outer and inner sidewall enclosures of said upper and lower structures is composed of generally planar pairs of vertical wall portions rigidly interconnected together at their opposite vertical edges. 12. In a nuclear fuel assembly including a bottom nozzle, a plurality of guide thimbles having upper and lower ends and being attached at said lower ends to said bottom nozzle and extending upwardly therefrom, an array of upstanding fuel rods extending along and spaced from said guide thimbles and spaced at their lower ends above said bottom nozzle, and a plurality of support grids axially spaced along and connected to said guide thimbles for supporting said array of upstanding fuel rods, a double enclosure top nozzle subassembly which permits increased fuel rod thermal growth and burnup, said top nozzle subassembly comprising: 13. The top nozzle subassembly as recited in claim 12, wherein said interengaging means includes: 14. The top nozzle subassembly as recited in claim 12, wherein said top plate of said upper structure includes an annular body having an inner peripheral edge defining a large central opening and aligned above said outer perimeter of said guide thimbles. 15. The top nozzle subassembly as recited in claim 16, where said outer peripheral edge of said top plate is defined by a rim projecting outwardly from an upper peripheral edge of said annular body so as to define an annular cavity surrounding said annular body for receiving an upper peripheral edge of said outer sidewall enclosure. 16. The top nozzle subassembly as recited in claim 14, wherein said top plate includes a plurality of indentations defined along the periphery and underside of said annular body in spaced relation to one another, said indentations facing outwardly of said body and downwardly therefrom, the portions of said annular body between said indentations forming downwardly protruding tabs along the periphery of said body for attaching said top plate to an upper peripheral edge of said outer sidewall enclosure. 17. The top nozzle subassembly as recited in claim 16, wherein said top plate further includes a plurality of pins inserted through said upper peripheral edge of said outer sidewall enclosure and said tabs of said annular body for securing said outer sidewall enclosure and said annular body together. 18. The top nozzle subassembly as recited in claim 16, wherein said inner sidewall enclosure of said lower structure has a plurality of upper edge portions spaced apart by notches defined between said upper edge portions, said upper edge portions and said notches being capable of mating respectively with said spaced indentations and tabs in said annular body of said top plate of said upper structure when said top nozzle subassembly is in said compressed condition. 19. The top nozzle subassembly as recited in claim 12, wherein said top plate of said upper structure and said lower adapter plate of said lower structure are generally rectangular in configuration having four sides defining four corners. 20. The top nozzle subassembly as recited in claim 19, wherein said biasing devices are a plurality of leaf springs disposed between said top plate and said lower adapter plate and movable between expanded and compressed states. 21. The top nozzle subassembly as recited in claim 20, wherein: 22. The top nozzle subassembly as recited in claim 12, wherein each of said outer and inner sidewall enclosures of said upper and lower structures is composed of generally planar pairs of vertical wall portions rigidly interconnected together at their opposite vertical edges. |
description | The present application claims the benefit of U.S. provisional patent application No. 61/488,940, filed May 23, 2011, by inventors Tomas Plettner et al., the disclosure of which is hereby incorporated by reference in its entirety. 1. Field of the Invention The present invention relates to technology for forming an electrical path through an insulating layer. 2. Description of the Background Art It is often desirable to electrically ground a substrate through an insulating layer on its surface. The substrate may be, for example, a silicon wafer or other semiconductor substrate. The insulating layer may be, for example, an oxide or nitride layer on a surface of the substrate. For example, in an electron beam inspection apparatus, a mechanism is typically used to ground a silicon wafer being inspected through its backside, where the backside is the side away from the integrated circuitry being manufactured. The conventional mechanism presses sharp pins of a hard metal against the insulating layer on the wafer backside to force electrical conduction paths from the pins to the bulk silicon wafer by either mechanical destruction of the insulating layer under the pins, or electrical arcing from the pins through the insulating layer, or a combination of both. Unfortunately, this conventional mechanism causes irreversible damage (electrical and/or mechanical) to the wafer backside. For example, an oxide layer on the backside of a silicon wafer may be scratched, or otherwise mechanically damaged by the sharp grounding pins, so as to expose the bulk silicon. Subsequently, as the wafer continues in the fabrication process, an etch process may be applied that inadvertently and undesirably etches the exposed silicon on the backside. In addition, if the oxide layer is very thick, the mechanical damage may be substantial enough to cause debris particles near the damaged area that may introduce misalignment of the wafer. One embodiment disclosed relates to an apparatus forming an electrical conduction path through an insulating layer on a surface of a substrate. In the case where charging of the substrate in question can be either negative or positive, two contacts with their respective radiation sources and with opposite bias polarities may be provided in the apparatus to allow for draining of either positive or negative charging. The first radiation source is configured to emit radiation to a first region of the insulating layer, and a first electrical contact is configured to apply a first bias voltage to the first region in a way that can drain the charge that is being deposited on the substrate. The second radiation source is configured to emit radiation to a second region of the insulating layer, and a second electrical contact is configured to apply a second bias voltage to the second region. The conductivities of the first and second regions are increased by the radiation such that conductive paths are formed through the insulating layer at those regions. In one implementation, the apparatus may be part of a wafer carrier and may be used in an electron beam imaging instrument. Another embodiment relates to a method of forming an electrical conduction path through an insulating layer. Radiation is emitted to a first region of the insulating layer, and a first bias voltage is applied to the first region. Radiation is emitted to a second region of the insulating layer, and a second bias voltage is applied to the second region. The conductivities of the first and second regions are increased by the radiation such that conductive paths are formed through the insulating layer at those regions. In one implementation, the method may be used in an electron beam imaging instrument to ground the substrate through the insulating layer. Other embodiments, aspects and features are also disclosed. FIG. 1 is a cross-sectional diagram of an apparatus 100 to form an electrical conduction path through an insulating layer in accordance with an embodiment of the invention. The apparatus 100 may be used to provide a mechanism that carries electrical current across an insulating layer (or layers) 104 that may range from less than one nanometer to several microns in thickness. Advantageously, mechanical damage to, and electrical breakdown of, the insulating layer may be avoided using the apparatus 100. As depicted in FIG. 1, the insulating layer 104 may be, for example, a layer of oxide and/or nitride formed on the backside of a semiconductor wafer 102. The apparatus 100 may also be employed to form an electrical conduction path through other insulating layers. The backside 104 of the wafer 102 may rest on a wafer carrier 106. As indicated in FIG. 1, the apparatus 100 may be utilized in an electron beam (e-beam) imaging instrument, such as, for example, an automated e-beam inspection instrument. Such an e-beam imaging instrument may focus an incident electron beam 108 onto the front-side of the wafer 102. The electron beam 108 comprises a current of negative charges to the bulk of the wafer 102. This negative charge current may build up an unwanted charge in the wafer 102. As described herein, the apparatus 100 of FIG. 1 provides a non-destructive mechanism for draining the unwanted charge build-up in the wafer 102. In accordance with an embodiment of the invention, in order to provide an electrical conduction path to drain the unwanted charge build-up without damaging the insulating layer, one or more electrical contacts 114 and one or more radiation sources 110 may be utilized. In the exemplary implementation shown in FIG. 1, there are two flexible mesh electrical contacts 114 that contact the insulating layer 104 of the wafer 102 and two radiation sources 110 in the vicinity of the contacts 114. As shown, each radiation source 110 may be arranged to emit radiation 111 through the openings of a corresponding contact mesh 114. Bias voltages (Bias1 115 and Bias 2 116) are applied to the contact meshes 114 by way of a conductive element 112. The conductive element 112 may, but does not need to, be part of a unit that surrounds the radiation source 110. The insulating layer 104 generally has a threshold electric field above which it breaks down. This threshold electric field may be referred to as the electrical breakdown field and depends on the material characteristics of the insulating layer 104. In accordance with an embodiment of the invention, each bias voltage (115 and 116) is preferably kept below the voltage which would cause electrical breakdown of the insulating layer 104. In other words, the electric fields caused by the bias voltages are preferably kept below the electrical breakdown field. The radiation sources 110 are chosen and arranged to add energy to the regions of the insulating layer 104 which are affected by the electric fields caused by the bias voltages (115 and 116). The radiation sources 110 may be advantageously configured to emit radiation 111 that serves to promote electrons into the conduction band, and/or holes into the valence band. In one embodiment, the radiation sources 110 may directly create electron-hole pairs within the insulating layer 104 by adding ionizing radiation. The ionizing radiation may be in the form of alpha particles, ion beams, high-energy photons, or electron beams. In a preferred embodiment, the ionizing radiation may be alpha particles emitted using a radioactive americium (Am) source. In another embodiment, the radiation sources 110 may cause photon-assisted injection of charge from the interface of the insulating layer 104 into the bulk of the insulating layer 104. Depending on the photon's energy that is employed, the charge may be either photo-emitted or tunneled into the insulating layer's conduction band. In certain embodiments, photons at ultraviolet (UV) or deep UV wavelengths may be used. In yet another embodiment, thermal injection of charge into the conduction band may be utilized. Thermal injection has been observed to be effective with nitride layers, for example. FIG. 2 is an electrical grounding diagram 200 of the apparatus 100 in accordance with an embodiment of the invention. A first bias voltage of +V1 may be applied by a first voltage source through a first protective resistor Rprotective to a first grid contact 114, and a second bias voltage of −V2 may be applied by a second voltage source through a second protective resistor Rprotective to a second grid contact 114. The protective resistors may be selected to have sufficient resistance to limit the maximum current flowing between the two grid contacts 114 so as to avoid damage to the voltage sources. Radiation 111 is emitted so that it impinges upon the insulating layer 104 through the grid contacts 114. The radiation 111 advantageously causes the promotion of electrons into the conduction band, and/or holes into the valence band, in the regions of the insulating layer 104 which is under the electric fields. As a result, a conductive path for electronic current is created through the insulating layer 104 in the region above each of the grid contacts 114. A first electrical current I1 may then flow through the first grid contact 114, and a second electrical current I2 may then flow through the second grid contact 114. As shown in the electrical grounding diagram 200, the first electrical current (the “source” current) I1=Ino—beam which flows between the first and second grid contacts 114, and the second electrical current (the “drain” current) I2=Ino—beam+Icharging, where Icharging is the electrical current due to charging from the electron beam 108 impinging upon the wafer 102. From Kirchhoff's law, Icharging=Ibeam−Iscattered, where Ibeam is the electron beam current and Iscattered is the scattered electron current. Conceptually, the first grid contact 114 with the positive voltage bias (+V1) sources electrons, while the second grid contact 114 with the negative voltage bias (−V2) drains electrons. The potential drop across the insulating layer 104 is relatively large, but of opposite signs for each contact. The net potential (V1+V2) is relatively small and may be adjusted to be zero. As shown, there is current (Ino—beam) that flows through the wafer 102 even in the absence of the electron beam 108 (i.e. even with Ibeam=0). When the electron beam is present, the addition of the charging current (Icharging) offsets the wafer bias, so a key parameter is the slope of the current versus bias (i.e. the small signal impedance). The beam current induced offset may be “re-adjusted” to zero (i.e. canceled) by applying a bias offset such that V1 is not equal to V2. FIG. 3 shows an equivalent circuit 300 in accordance with an embodiment of the invention. As shown, a first impedance Z1 is effectively present between the first voltage source at the voltage +V1 (the “source” voltage) and the bulk of the wafer 102 at voltage VW, and a first impedance Z2 is effectively present between the bulk of the wafer 102 at voltage VW and the second voltage source at the voltage −V2 (the “drain” voltage). Z1 is approximately equal to the small-signal impedance value for the positively-biased region of the insulating layer 104, and Z2 is approximately equal to the small-signal impedance value for the negatively-based region of the insulating layer 104. From Ohm's law, ΔV1=V1−VW=I1Z1, and ΔV2=VW−(−V2)=VW+V2=I2Z2. As discussed above in relation to FIG. 2, I1=Ino—beam, and I2=Ino—beam+Icharging. FIG. 4 is a flow chart of a method 400 for selecting and applying bias voltages of the backside contacts in accordance with an embodiment of the invention. Positive bias current-voltage (IV) characteristics are found 402 for the insulating layer 104, and a determination may be made 406 from the positive I-V characteristics as to the positive bias 115 (+V1) to apply to the first mesh contact 114 so as to achieve a target current for Ino—beam. In addition, negative bias IV characteristics are found 404 for the insulating layer 104, and a determination may be made 408 from the negative I-V characteristics as to the negative bias 116 (−V2) to apply to the second mesh contact 114 so as to achieve a target current for Ino—beam. The positive bias 115 (+V1) is applied to the first contact 114, and the negative bias 116 (−V2) is applied to the second contact 114 so that the targeted current of Ino—beam should flow in the absence of an electron beam 108. If the wafer 102 is within an e-beam instrument, then the e-beam 108 may be focused 414 onto the surface of the wafer 102. The charging current Icharging due to the beam 108 is then drained 416 as part of the second electrical I2 through the second contact 114. This method 400 advantageously grounds the wafer 102 to avoid charge build-up in a non-destructive manner. Unlike conventional grounding techniques which employ a ground (zero) potential at the electrical pin contacts, the method 400 disclosed herein employs non-zero potentials at the electrical contacts to achieve an essentially neutral (zero or near zero) potential in the wafer bulk. FIG. 5 shows a positive bias current-voltage characteristic curve 500 for an example insulating layer in accordance with an embodiment of the invention, and FIG. 6 shows a negative bias current-voltage characteristic curve for the example insulating layer in accordance with an embodiment of the invention. The example insulating layer is a nitride layer 704 on a backside of a silicon wafer 702 as depicted in FIG. 7. As shown in FIGS. 5 and 6, an electrical current of 0.7 microamperes (μA) may be targeted. From the I-V curves in FIGS. 5 and 6, it is seen that this target current is reached at a positive bias of about +53 volts and a negative bias of about −57 volts. In accordance with steps 410 and 412 in FIG. 4, these bias voltages may be applied, respectively, to the first and second contacts 706 as depicted in FIG. 7. FIG. 8 is a cross-sectional diagram of an alternate apparatus 800 to form an electrical conduction path through an insulating layer in accordance with an embodiment of the invention. The apparatus 800 in FIG. 8 is similar to the apparatus 100 described above in relation to FIG. 1. However, while the apparatus 100 in FIG. 1 has two electrical contacts 114 and two radiation sources 110, the apparatus 800 in FIG. 8 has only one electrical contact 114 and one radiation source 110. The apparatus 800 in FIG. 8 may be applied in cases where there is only one type of charging (i.e. only positive or only negative charging) to be drained. In contrast, the apparatus 100 in FIG. 1 has flexibility to drain either sign charge. For example, in electron beam instruments with variable landing energy, the charging may be of either sign. The above-described diagrams are not necessarily to scale and are intended be illustrative and not limiting to a particular implementation. In the above description, numerous specific details are given to provide a thorough understanding of embodiments of the invention. However, the above description of illustrated embodiments of the invention is not intended to be exhaustive or to limit the invention to the precise forms disclosed. One skilled in the relevant art will recognize that the invention can be practiced without one or more of the specific details, or with other methods, components, etc. In other instances, well-known structures or operations are not shown or described in detail to avoid obscuring aspects of the invention. While specific embodiments of, and examples for, the invention are described herein for illustrative purposes, various equivalent modifications are possible within the scope of the invention, as those skilled in the relevant art will recognize. These modifications can be made to the invention in light of the above detailed description. The terms used in the following claims should not be construed to limit the invention to the specific embodiments disclosed in the specification and the claims. Rather, the scope of the invention is to be determined by the following claims, which are to be construed in accordance with established doctrines of claim interpretation. |
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abstract | A nuclear power plant control system (3) is provided with detection units (30a to 30d) which detect phenomena that occurs in a nuclear power plant for each of four systems, a trip control device (20) which starts, in the case where a signal that indicates an occurrence of the phenomenon is input from at least a predetermined number of signal lines out of signal lines of two systems, processing corresponding to the phenomenon, and majority circuits (50a and 50b) which are provided for each signal line of the two systems and each output, in the case where the phenomenon is detected by N or more detection units out of the detection units (30a to 30d), a signal that indicates an occurrence of the phenomenon to a corresponding signal line. |
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abstract | A system for passively cooling nuclear fuel in a pressurized water reactor during refueling that employs gravity and alignment of valves using battery reserves or fail in a safe position configurations to maintain the water above the reactor core during reactor disassembly and refueling. A large reserve of water is maintained above the elevation of and in fluid communication with the spent fuel pool and is used to remove decay heat from the reactor core after the reaction within the core has been successfully stopped. Decay heat is removed by boiling this large reserve of water, which will enable the plant to maintain a safe shutdown condition without outside support for many days. |
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description | The invention relates to the use of a mixture comprising erbium and praseodymium as a radiation attenuating composition, i.e. as a composition having the property of attenuating ionising radiation, in particular X- and gamma-type electromagnetic radiation. It also relates to a radiation attenuating material comprising a radiation attenuating composition comprising erbium and praseodymium, as well as a protective article which provides individual or group protection against ionising radiation and comprises said material. The invention finds application in all fields in which protection against ionising radiation may be sought and, in particular, in the fields of nuclear medicine (scintigraphy, radiotherapy, etc.), radiology, medical imaging, and the nuclear industry. In a certain number of professions, it is normal to use clothing and other articles to protect against ionising radiation. This is particularly the case in the fields of medicine, radiology, or medical imaging, where ionising radiation is used for diagnostic and therapeutic purposes. It is also the case in the plastic materials industry where irradiations are used to obtain chemical effects of polymerisation, grafting, cross-linking or degradation of polymers; in the nuclear industry, where operators are exposed to a risk of irradiation, particularly during the handling of powders of nuclear fuels or from the dismantling of facilities; or in inspection and control laboratories, for example of manufactured parts, where analytical techniques based on the use of ionising radiation are employed. Most radiation protection articles currently available on the market comprise a matrix, the nature of which depends on the destination of said articles and which contain lead, either in the form of sheets, or in the form of fine particles, the lead then being able to be in the metal, oxide or salt state. Given the toxicity of lead and compounds thereof, the manufacture of such protective articles requires heavy and costly equipment to prevent any contamination of the personnel in charge of this manufacture. In addition, the elimination of waste from the manufacture of these articles as well as that of protective articles after use requires specific collection and treatment channels, failing which they are quite simply disposed of in discharges with all the harmful consequences on the environment which that can imply. Also, it has recently been proposed to replace the use of lead as radiation attenuating agent by that of other metals which are also capable of attenuating ionising radiation but which are not toxic or, in any case, have lower toxicity than that of lead. Thus, for example, PCT international application WO 2006/069007 [1] advocates using a radiation attenuating composition composed of a salt of elementary barium, tungsten and bismuth. Patent application US 2008/0128658 [2] describes the use of a composition comprising the oxide of gadolinium Gd2O3, tungsten and one or more oxides of rare earths other than gadolinium, such as LaO3, CeO2, Nd2O3, Pr6O11, Eu2O3 and Sm2O3. Patent application FR 2 948 672 [3] advocates the use of a composition composed of oxides of tungsten, bismuth and lanthanum. PCT international application WO 2005/017556 [4] proposes using a composition comprising at least two elements selected from antimony, bismuth, iodine, tungsten, tin, tantalum, erbium, barium, salts, compounds and alloys thereof, whereas patent application DE 10 2006 958 [5] describes a multilayer radiation protection material, certain layers of which comprise a radiation attenuating element selected from tin, antimony, iodine, caesium, barium, lanthanum, cerium, praseodymium and neodymium, optionally associated with a second radiation attenuating element having, for its part, an atomic number ranging from 60 to 70. Although it cannot be contested that erbium and praseodymium form part of the chemical elements that are cited in the aforementioned references [2], [4] and [5] as being capable of being used in radiation attenuating compositions, it turns out that nothing is said in these references on the real capacities of these two elements, taken separately or in combination, to attenuate ionising radiation. Yet, it turns out that, within the scope of their works, the inventors have observed that a mixture comprising erbium or a compound thereof and praseodymium or a compound thereof has particularly interesting radiation attenuation properties, and that these properties may advantageously be harnessed to form materials and protective articles able to assure very efficient protection against ionising radiation, in particular X- and gamma-type electromagnetic radiation. It is on the basis of this observation that the invention is based. The subject-matter of the invention is thus, firstly, the use of a mixture comprising: 30 to 70% by mass of erbium or of a compound thereof; 20 to 50% by mass of praseodymium or of a compound thereof; and 0 to 50% by mass of bismuth or of a compound thereof; as a radiation attenuating composition. The basis of the principle of radiation attenuation implemented within the scope of the invention is an interaction that takes place between, on the one hand, the photons from an ionising radiation and, on the other hand, at least one radiation attenuating chemical element, the latter absorbing part of the energy of said photons. This ionising radiation may be a gamma-type electromagnetic radiation, when this is emitted by one or more radioactive atoms during their disintegration. This ionising radiation may also be an X-type electromagnetic radiation, when this is produced by an X-ray generator, within which a potential difference ranging usually from several tens to several hundreds of kilovolts (kV) is applied. The probability and the intensity of this interaction are closely linked to various parameters, such as the nature of the radiation attenuating chemical element, the binding forces between the atomic nucleus of said element and the different shells of its electron cloud, or the energy of the ionising radiation. In concrete terms, the capacity of a chemical element to attenuate radiation may be measured by a mass attenuation coefficient, which is proportional to this probability of interaction, this also being known as “cross-section”. Thus, the higher the cross-section the greater the attenuation. For a same element of the periodic table of elements, the cross-section exhibits discontinuities linked to the binding energies of the different electron shells of this element. The phenomenon of absorption of a photon (gamma or X) by the radiation attenuating chemical element is observed when the energy of the photon is substantially greater than the binding energy of one of the electrons of said chemical element. This phenomenon increases significantly when the energy of said photon is sufficiently high to expulse an electron from a deeper electron shell of the radiation attenuating chemical element. The inventors have thus been able to demonstrate, as is explained hereafter, the existence, for erbium and compounds thereof, of an absorption maximum for a photonic energy of the order of 60 kiloelectron-volts (keV). This absorption maximum is, moreover, greater than that measured for lead at the same energy. The interaction between the photons from the ionising radiation and the radiation attenuating chemical element, as we have described above, can take place according to several effects, such as the photoelectric effect, the Compton effect or the materialization effect. The preponderant effects are closely linked to the atomic number of the chemical element that undergoes the absorption, but also to the energy of the absorbed radiation. In the case of erbium, element of atomic number 68, subjected to an ionising radiation of 60 keV, the interaction mainly takes place according to the photoelectric effect, which signifies that each of the photons of the ionising radiation is absorbed while expelling an electron from one of the electron shells of the atom of erbium. This subsequently reorganizes the electron vacancy created, and restores the energy acquired by emitting one or more photons. Thus, for this element, these photons constitute the basis of an X-type secondary radiation, of energy mainly centred on 52 keV. The inventors have thus been able to demonstrate that erbium and compounds thereof, particularly oxides thereof, turn out to be particularly efficient in the radiation attenuation field, when they are subjected to an ionising radiation, for example an X- or gamma-type electromagnetic radiation, of energy mainly centred on 60 keV. Energy “mainly centred” on 60 keV is taken to mean an energy for which a proportion greater than or equal to 80% of the distribution of photons of an energy spectrum, which corresponds to this radiation, has an energy equal to 60 keV. This type of radiation may, for example, come from X-ray generators within which a potential difference, ranging for example from 80 to 150 kV, is applied. In particular, for potential differences of 80 and 140 kV, the inventors have in particular been able to demonstrate the existence of a high distribution of photons having an energy approximately equal to 60 keV. This type of radiation may further be the main radiation emitted by a nuclear fuel, for example MOX (constituted of a mixture of oxides of plutonium and uranium), for which this main radiation corresponds to the emission of a gamma photon by americium-241, obtained itself by β− disintegration of radioactive plutonium-241. The existence of an X-type secondary electromagnetic radiation, as described previously, has also been taken into consideration by the inventors. Consequently, and according to the invention, the erbium or the erbium compound is used, in the radiation attenuating composition, in combination with praseodymium or a compound thereof. In fact, by using a radiation attenuating composition associating erbium or a compound thereof with praseodymium or a compound thereof, the inventors have thus been able to demonstrate, as will be shown hereafter, the existence of two absorption maxima: thanks to the erbium or to the compound thereof, for example sesquioxide of erbium(III), an absorption maximum for a photonic energy of the order of 60 keV; and thanks to the praseodymium or to the compound thereof, for example oxide of praseodymium(III-IV), another absorption maximum for a photonic energy of the order of 45 keV, corresponding to the energy of the X-type secondary radiation emitted by erbium, which has been described previously. The erbium compound is, preferably, an erbium oxide and, even more preferably, sesquioxide of erbium(III), of formula Er2O3, whereas the praseodymium compound is, preferably, a praseodymium oxide and, even more preferably, an oxide selected from oxide of praseodymium(III), oxide of praseodymium(IV) and oxide of praseodymium(III-IV), of respective formulas Pr2O3, PrO2 and Pr6O11. Oxide of praseodymium(III-IV) is quite particularly preferred. When the radiation attenuating composition according to the invention comprises such oxides of erbium and of praseodymium, it comprises, preferably, 55 to 65% by mass of erbium oxide and 35 to 45% by mass of praseodymium oxide; better still, the radiation attenuating composition comprises (60±2) % by mass of erbium oxide and (40±2) % by mass of praseodymium oxide. Furthermore, the inventors have also been able to show that the protection spectrum conferred by a radiation attenuating composition, which comprises erbium or a compound thereof and praseodymium or a compound thereof, may be further widened by using them jointly with bismuth or a compound thereof. Also, according to a particularly preferred disposition of the invention, the erbium or the erbium compound and the praseodymium or the praseodymium compound are used within the radiation attenuating composition, jointly with at least bismuth, introduced in elementary form or in the form of a compound, for example the sesquioxide of bismuth(III), of formula Bi2O3, in proportions that depend in particular on the energy of the ionising radiation received by the radiation attenuating composition thereby constituted. Thus, by using a radiation attenuating composition associating erbium or a compound thereof, praseodymium or a compound thereof and bismuth or a compound thereof, the inventors have been able to demonstrate, as will be shown hereafter, the existence of three absorption maxima: thanks to the erbium or to the erbium compound, for example sesquioxide of erbium(III), an absorption maximum for a photonic energy of the order of 60 key; thanks to the praseodymium or to the praseodymium compound, for example oxide of praseodymium(III-IV), an absorption maximum for a photonic energy of the order of 45 key; finally, thanks to the bismuth or to the bismuth compound, an absorption maximum for a photonic energy of the order of 90 keV, to which very satisfactory radiation attenuation properties, for ionising radiation having photonic energies of the order of 40 keV and less, are added. Moreover, it may be noted that the use of a composition associating erbium or a compound thereof, praseodymium or a compound thereof and bismuth or a compound thereof enables the attenuation of an ionising radiation having a wide energy range, for example comprised between 0 and 100 keV, the radiation attenuation properties of each of said three elements being not discrete but continuous. Preferably, the bismuth is used in elementary form. Also preferably, when bismuth is present in the radiation attenuating composition, the latter comprises 30 to 45% by mass of erbium oxide, 20 to 30% by mass of praseodymium oxide and 30 to 45% by mass of bismuth; better still, it comprises 33 to 42% and, in a particularly preferred manner, (36±2) % by mass of erbium oxide, 22 to 28% and, in a particularly preferred manner, (24±2) % by mass of praseodymium oxide, and 30 to 45% and, in a particularly preferred manner, (40±2) % by mass of bismuth. In a variant, it is also possible to associate erbium or the compound thereof and praseodymium or the compound thereof with antimony, barium, tin, tantalum, tungsten, uranium, one of their compounds and mixtures thereof. According to the invention, the erbium or compound thereof, the praseodymium or compound thereof and, if need be, the bismuth or compound thereof are, preferably, used in the form of powders dispersed in a matrix. The subject-matter of the invention is thus also a radiation attenuating material that comprises a matrix in which a radiation attenuating composition is dispersed, the composition being in the form of a powder, and which is characterised in that said composition comprises: 30 to 70% by mass of erbium or of a compound thereof; 20 to 50% by mass of praseodymium or of a compound thereof; and 0 to 50% by mass of bismuth or of a compound thereof. As mentioned previously, the erbium compound is typically an oxide and, in particular, the sesquioxide of erbium(III), of formula Er2O3. Similarly, the praseodymium compound is typically an oxide, which is, preferably, selected from oxide of praseodymium(III), oxide of praseodymium(IV) and oxide of praseodymium(III-IV), of respective formulas Pr2O3, PrO2 and Pr6O11, the oxide of praseodymium(III-IV) being quite particularly preferred. When the radiation attenuating composition according to the invention comprises such oxides of erbium and of praseodymium, it comprises, preferably, 55 to 65% by mass of erbium oxide and 35 to 45% by mass of praseodymium oxide; better still, this composition comprises (60±2) % by mass of erbium oxide and (40±2) % by mass of praseodymium oxide. When the radiation attenuating composition according to the invention comprises an erbium oxide, a praseodymium oxide and bismuth, it comprises, preferably, 30 to 45% by mass of erbium oxide, 20 to 30% by mass of praseodymium oxide and 30 to 45% by mass of bismuth; better still, it comprises 33 to 42% and, in a particularly preferred manner, (36±2) % by mass of erbium oxide, 22 to 28% and, in a particularly preferred manner, (24±2) % by mass of praseodymium oxide, and 30 to 45% and, in a particularly preferred manner, (40±2) % by mass of bismuth. According to the invention, the respective proportions of the matrix and of the radiation attenuating composition in the material can vary to a large extent as a function of the use for which said material is intended and, in particular, the level of radiation attenuation sought within the context of said use. This being so, it is generally preferred that the matrix represents 10 to 25% by mass of the mass of the material and that the radiation attenuating composition represents, for its part, 75 to 90% by mass of the mass of the material. For the manufacture of radiation protection articles and, in particular, individual protective articles such as a protective overall, it is preferred that the matrix represents (15±2) % by mass of the mass of the material and that the radiation attenuating composition represents (85±2) % by mass of the mass of the material. Furthermore, and so as to obtain a distribution of this composition that is the most homogeneous possible in the matrix, the radiation attenuating composition is, preferably, constituted of particles of which at least 90% by number have an average particle size less than or equal to 20 μm and, better still, less than or equal to 1 μm. As for the matrix, it is also chosen as a function of the use for which the radiation attenuating material is intended. Thus, for example, for the manufacture of an individual protective article of the type glove, overall, chasuble, jacket, skirt, oversleeve, thyroid protector, gonad protector, armpit protective clothing, ocular protection headband, operative field, curtain, sheet, the desired mechanical properties, the characteristics of flexibility and comfort of this article are oriented preferably towards a matrix based on a thermoplastic material, in particular, polyvinyl chloride, or based on an elastomeric material, selected in particular from natural rubber, synthetic polyisoprenes, polybutadienes, polychloroprenes, chlorosulphonated polyethylenes, polyurethane elastomers, fluorinated elastomers (or fluoroelastomers), isoprene-iso-butylene copolymers (or butyl rubbers), copolymers of ethylene-propylene-diene (or EPDM), sequenced copolymers of styrene-isoprene-styrene (or SIS), sequenced copolymers of styrene-ethylene-butylene-styrene (or SEBS), and mixtures thereof. In a variant, for the manufacture of a group protective article of the type bedding, panel, protective screen, the search for characteristics of durability and resistance to wear of material leads preferably towards matrices of silicious type, in particular glass, matrices based on a thermosetting resin, selected in particular from resins of type epoxides, vinyl esters and unsaturated polyesters, or instead a material based on a thermoplastic, selected in particular from polyethylene, polypropylene, a polycarbonate, for example, bisphenol A polycarbonate, acrylonitrile-butadiene-styrene (or ABS) and products obtained by co-extrusion of ABS with compounds of (meth)acrylate type, such as polymethylmethacrylate (or PMMA). The subject-matter of the invention is also an article providing protection against ionising radiation, comprising a radiation attenuating material as defined previously. Preferably, the protective article is an individual protective article such as a glove, an overall, a chasuble, a jacket, a skirt, an oversleeve, a thyroid protector, a gonad protector, an armpit protective clothing, an ocular protection headband, an operating field, a curtain, a sheet, or a group protective article such as a bedding, a panel or a protective screen. The invention has numerous advantages. In fact, it makes it possible to produce materials and protective articles which have remarkable properties of attenuating ionising radiation, in particular X- and gamma-type electromagnetic radiation, of energy that can lie within a wide range, typically comprised between 0 and 100 keV, and does so, from metals and metal oxides which do not have any toxicity known to date for human health and the environment. Moreover, the elimination of the waste stemming from their manufacture thus does not require any specific collection and treatment channel. Finally, in a similar manner, the elimination of these materials and protective articles after use does not require any specific channel other than those that are imposed by a potential contamination by toxic or radioactive materials. Other characteristics and advantages of the invention will become clearer on reading the complement of description that follows, which relates to examples of manufacture of materials according to the invention as well as a demonstration of the radiation attenuation properties of these materials. Obviously, these examples are only given by way of illustration of the subject-matter of the invention and do not in any way constitute a limitation of said subject-matter. Five samples, respectively E1, E2, E3, E4 and E5, of materials according to the invention were produced. The samples E1, E2 and E3 correspond to materials that comprise a radiation attenuating composition composed of Er2O3 and of Pr6O11 whereas the samples E4 and E5 correspond to materials that comprise a radiation attenuating composition composed of Er2O3, of Pr6O11, and of bismuth in elementary form. These samples, which are in the form of squares of approximately 30 centimeter sides, are produced by coating technique. Moreover, these samples implement a radiation attenuating composition in the form of powders of which at least 90% of the particles constituting said powders have an average particle size less than or equal to 20 μm. The characteristics specific to each of these samples are grouped together in Table 1. TABLE 1SampleE1E2E3E4E5Thickness4.62.35.21.63.2(mm)Basis weight13.4 5.813.8 4.89.6(kg/m2)Base of the matrixSiliconePVCPVCSiliconeSiliconeMass proportion75/2568/3268/3275/2575/25composition/matrix(%/%)Mass proportion60/40/070/30/070/30/036/24/4036/24/40Er2O3/Pr6O11/Bi in thecomposition (%/%/%) The samples obtained in Example 1 above were subjected to tests intended to evaluate their capacity to attenuate X-type ionising radiation, which comes from X-ray generators within which a particular potential difference is applied, or of gamma-type, which are for example emitted by powders entering into the manufacture of nuclear fuels. 1. Radiation Attenuation Properties in the Presence of an X-Type Ionising Radiation The properties of attenuation of an X-type ionising radiation by materials according to the invention are evaluated by applying the provisions of the NF EN 61331-1 standard, entitled “Protective devices against diagnostic medical X-radiation. —Part 1: Determination of attenuation properties of materials”. The results as obtained with diverse potential differences are expressed in terms of theoretical lead equivalent thickness, noted etheo(X), and of measured lead equivalent thickness, noted eexp(X). A gain factor is also defined, noted FX, for a potential difference and particular weight proportions of Er2O3/Pr6O11/Bi within the radiation attenuating composition, as being the ratio of eexp(X) to etheo(X). When the ratio FX equals 1, the efficiency of a material is equivalent, in radiation attenuation terms, to that of a material of same basis weight but constituted uniquely of lead. The results obtained for the samples E1, E2, E4 and E5 are shown in Table 2 below. TABLE 2MassproportionPotentialEr2O3/Pr6O11/Bietheo(X)eexp(X)FXdifference (kV)Sample(%/%/%)(mm)(mm)(Ø)80E160/40/00.881.351.53E270/30/00.350.431.22E436/24/400.310.431.37E536/24/400.631.031.63110E436/24/400.310.481.52E536/24/400.631.021.61150E270/30/00.350.401.14E436/24/400.310.391.24E536/24/400.630.761.19 Gain factors comprised between 1.14 and 1.63 are obtained with the materials according to the invention, which signifies that said materials have enhanced radiation attenuating properties compared to materials containing a radiation attenuating agent constituted uniquely of lead. 2. Radiation Attenuation Properties in the Presence of a Gamma-Type Ionising Radiation The properties of attenuation of a gamma-type ionising radiation by materials according to the invention are evaluated by means of a device implementing said materials, placed at a certain distance between, on the one hand, a radioactive source constituted of americium-241, which emits a gamma-type ionising radiation of 59 keV energy, and on the other hand, a spectrometer on which is assembled a germanium gamma detector. The method employed consists in determining the attenuation of the gamma-type radiation from americium-241, by measuring the surface of the photoelectric absorption peaks recorded by the detector. This surface is compared, by the same method, to surfaces obtained with lead screens of known thickness. As in the preceding paragraph 1, a theoretical lead equivalent thickness, noted etheo(γ), is defined and calculated from the basis weight of the materials tested, and from the density of lead in metal form. In other words, this thickness corresponds to the thickness of a material of same weight as the materials tested, but composed uniquely of lead. A measured lead equivalent thickness, noted eexp(γ), is again defined. A gain factor Fγ, corresponding to the ratio eexp(γ)/etheo(γ), is also defined. The results obtained for the samples E2 and E3 are shown in the Table 3 below. TABLE 3MassproportionEr2O3/Pr6O11/Bietheo(X)eexp(X)FYSample(%/%/%)(mm)(mm)(Ø)E270/30/00.350.802.28E370/30/00.821.672.03 Gain factors greater than 2 are obtained with the materials according to the invention, which thus have enhanced radiation attenuating properties compared to materials containing a radiation attenuating agent uniquely constituted of lead. A graphical representation of the cross-section, noted n, as a function of the photonic energy, noted E, is shown in FIG. 7. The thick line curve, which represents the cross-section of photons from a gamma-type ionising radiation emitted by americium-241, as a function of the photonic energy, has a maximum corresponding to a high distribution of photons having an energy mainly centred on 59.6 keV. By comparing the surface portions situated under the thin line curve, a strong attenuation of the radiation of energy mainly centred on 59.6 keV is observed. Moreover, it is also possible to observe the emission of a secondary X-type radiation, which is materialized in the form of two rays noted “RS” and “RS′” in FIG. 7, and the respective energies of which are mainly centred on 49 and 55 keV. As previously exposed, such a material according to the invention may be used for purposes of attenuation of radiation from MOX fuel. In this respect, and as a complement, it may be added that, depending on the variability of the isotopic composition of this fuel, this being placed at a short distance from a measuring point, typically 50 centimeters, this gamma-type ionising radiation represents a proportion ranging from 75 to 85% of all the gamma- and X-radiation from the latter. This high proportion makes all the more legitimate the implementation of a radiation attenuating composition as described above in the manufacture of protective articles against ionising radiation. [1] International application PCT WO 2006/069007 [2] Patent application US 2008/0128658 [3] Patent application FR 2 948 672 [4] International application PCT WO 2005/017556 [5] Patent application DE 10 2006 958 |
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050248030 | claims | 1. An emergency device for control of displacement of a conveying carrier for a nuclear fuel assembly, in which said conveying carrier moves under the effect of a push-pull chain actuated by a driving mechanism selectively causing winding of said chain on itself in spiral so as to pull said carrier, and unwinding of said chain by forming a rigid carrier pushing element, including a connecting member between said chain and said carrier, said connecting member being formed of two portions connected to each other by a pin adapted to be broken when required for disconnecting the two portions and to free the carrier from the chain, wherein the breaking of said pin is due to the effect of traction exerted remotely from said carrier on a control cable, such that, after breaking, the cable can apply a traction on said carrier so as to bring it back freely, independently of said chain, to a predetermined position, the connecting member between said chain and said carrier including a fastening member attached to said carrier and connected to two parallel flanges of a link of said chain by means of a transverse spindle, said fastening member and said spindle being connected by said pin, said pin being parallel to said flanges. 2. An emergency device according to claim 1, wherein the transverse spindle extends freely through passages formed in alignment in said flanges and said fastening member, respectively. 3. An emergency device according to claim 1, wherein said transverse spindle is slidably mounted in a first housing of a support block, carried by said carrier and including a rack actuated by a flat pinion mounted so as to rotate freely on an axis perpendicular to the pinion and rigidly connected to the support block. 4. An emergency device according to claim 3, wherein said pinion actuating said rack of said spindle is driven in rotation by a second spindle mounted in the support block inside a second housing located in a plane of said first housing but perpendicularly to it, said second spindle also including a rack in mesh with the pinion in such manner that the rotation of the latter causes simultaneous and opposite displacement of the spindles, respectively toward the inside and the outside of the support block. 5. An emergency device according to claim 4, wherein said second spindle is rigidly connected at its end outside said support block to a member connecting with said traction cable. 6. An emergency device according to claim 5, wherein said connecting member includes a whorl adapted to bear at the end of the stroke of the second spindle under the effect of the pinion on an abutment rigidly connected to said carrier, in such manner that a traction effort exerted on said cable once said pin is broken brings said whorl against the abutment, thereby causing entrainment of said carrier by said cable. 7. An emergency device according to claim 3, wherein said support block of said pinion is mounted inside a tight casing which is supported by said axis of rotation pinion, said pinion being in turn disposed underneath a lower face of said carrier. 8. An emergency device according to claim 1, wherein said control cable is actuated by a winch located at a distance from said carrier and on a drum on which is wound said cable, guided toward said carrier by return pulleys. 9. An emergency device according to claim 8, wherein said cable winds itself on a mobile pulley provided with a counterweight exerting constant tension on said cable, which remains permanently taut in all positions of said carrier. 10. An emergency device according to claim 1, wherein said carrier includes wheels moving on rails extending parallel to a working plane of said push-pull chain. 11. An emergency device according to claim 10, wherein said chain winds itself on a toothed control driven by a return mechanism from a motor reduction unit. |
claims | 1. An ultraviolet irradiation device for irradiating with ultraviolet rays an ultraviolet curable protective tape joined to a surface of a semiconductor wafer, comprising:a holding table to mount and hold the semiconductor wafer with the protective tape joined thereto;a drive mechanism to turn the holding table;ultraviolet light emitting diodes arranged at least in a radial direction of the semiconductor wafer; anda control unit to maintain a uniform accumulated quantity of light in an area of the protective tape where ultraviolet rays are applied upon irradiation of a surface of the protective tape with ultraviolet rays from the ultraviolet light emitting diodes while turning the holding table that places and mounts the semiconductor wafer with the protective tape joined thereto,wherein supply voltage is controlled such that ultraviolet intensity from the ultraviolet light emitting diodes increases outwardly of the semiconductor wafer. 2. The ultraviolet irradiation device according to claim 1, further comprising:an amplifier to amplify the supply voltage to the ultraviolet light emitting diodes; anda controller to control the amplifier. 3. An ultraviolet irradiation device for irradiating with ultraviolet rays an ultraviolet curable protective tape joined to a surface of a semiconductor wafer, comprising:a holding table to mount and hold the semiconductor wafer with the protective tape joined thereto;a drive mechanism to turn the holding table;ultraviolet light emitting diodes arranged at least in a radial direction of the semiconductor wafer; anda control unit to maintain a uniform accumulated quantity of light in an area of the protective tape where ultraviolet rays are applied upon irradiation of a surface of the protective tape with ultraviolet rays from the ultraviolet light emitting diodes while turning the holding table that places and mounts the semiconductor wafer with the protective tape joined thereto,wherein a distance from the surface of the protective tape to the ultraviolet light emitting diodes decreases outwardly of the semiconductor wafer. 4. The ultraviolet irradiation device according to claim 3, further comprising:a sensor to measure intensity of ultraviolet rays from the ultraviolet light emitting diodes;a lifting mechanism to change each level of the ultraviolet light emitting diodes; anda level controller to control each level of the ultraviolet light emitting diodes based on detected result of the sensor through operation of the lifting mechanism so as to maintain a uniform accumulated quantity of light per area on the surface of the protective tape. 5. The ultraviolet irradiation device according to claim 4, further comprising:a drive mechanism to move the sensor between a position where the ultraviolet light emitting diodes apply ultraviolet rays and a position where the ultraviolet light emitting diodes apply no ultraviolet rays. 6. An ultraviolet irradiation device for irradiating with ultraviolet rays an ultraviolet curable protective tape joined to a surface of a semiconductor wafer, comprising:a holding table to mount and hold the semiconductor wafer with the protective tape joined thereto;a drive mechanism to turn the holding table;ultraviolet light emitting diodes arranged at least in a radial direction of the semiconductor wafer; anda control unit to maintain a uniform accumulated quantity of light in an area of the protective tape where ultraviolet rays are applied upon irradiation of a surface of the protective tape with ultraviolet rays from the ultraviolet light emitting diodes while turning the holding table that places and mounts the semiconductor wafer with the protective tape joined thereto,wherein spaces between the ultraviolet light emitting diodes are smaller outwardly of the semiconductor wafer. 7. The ultraviolet irradiation device according to claim 6, further comprising:a drive mechanism to move the ultraviolet light emitting diodes separately in a horizontal direction; anda controller to adjust the spaces between the ultraviolet light emitting diodes by control of the drive mechanism. 8. An ultraviolet irradiation device for irradiating with ultraviolet rays an ultraviolet curable protective tape joined to a surface of a semiconductor wafer, comprising:a holding table to mount and hold the semiconductor wafer with the protective tape joined thereto;a drive mechanism to turn the holding table;ultraviolet light emitting diodes arranged at least in a radial direction of the semiconductor wafer;a control unit to maintain a uniform accumulated quantity of light in an area of the protective tape where ultraviolet rays are applied upon irradiation of a surface of the protective tape with ultraviolet rays from the ultraviolet light emitting diodes while turning the holding table that places and mounts the semiconductor wafer with the protective tape joined thereto; anda lighting control unit to control intermittent lighting of each ultraviolet light emitting diode so as to extend a lighting duration of each ultraviolet light emitting diode outwardly of the semiconductor wafer. 9. An ultraviolet irradiation device for irradiating with ultraviolet rays an ultraviolet curable protective tape joined to a surface of a semiconductor wafer, comprising:a holding table to mount and hold the semiconductor wafer with the protective tape joined thereto;a drive mechanism to turn the holding table;ultraviolet light emitting diodes arranged at least in a radial direction of the semiconductor wafer;a control unit to maintain a uniform accumulated quantity of light in an area of the protective tape where ultraviolet rays are applied upon irradiation of a surface of the protective tape with ultraviolet rays from the ultraviolet light emitting diodes while turning the holding table that places and mounts the semiconductor wafer with the protective tape joined thereto; anda screen that is interposed between the semiconductor wafer and the ultraviolet light emitting diodes and has a slit formed therein that fans outwardly from the center of the semiconductor wafer. 10. An ultraviolet irradiation device for irradiating with ultraviolet rays an ultraviolet curable protective tape joined to a surface of a semiconductor wafer, comprising:a holding table to mount and hold the semiconductor wafer with the protective tape joined thereto;a drive mechanism to turn the holding table;ultraviolet light emitting diodes arranged at least in a radial direction of the semiconductor wafer;a control unit to maintain a uniform accumulated quantity of light in an area of the protective tape where ultraviolet rays are applied upon irradiation of a surface of the protective tape with ultraviolet rays from the ultraviolet light emitting diodes while turning the holding table that places and mounts the semiconductor wafer with the protective tape joined thereto; anda filter that is interposed between the semiconductor wafer and the ultraviolet light emitting diodes and has increased ultraviolet transmittance outwardly of the semiconductor wafer. 11. An ultraviolet irradiation device for irradiating with ultraviolet rays an ultraviolet curable protective tape joined to a surface of a semiconductor wafer, comprising:a holding table to mount and hold the semiconductor wafer with the protective tape joined thereto;a drive mechanism to turn the holding table;ultraviolet light emitting diodes arranged at least in a radial direction of the semiconductor wafer;a control unit to maintain a uniform accumulated quantity of light in an area of the protective tape where ultraviolet rays are applied upon irradiation of a surface of the protective tape with ultraviolet rays from the ultraviolet light emitting diodes while turning the holding table that places and mounts the semiconductor wafer with the protective tape joined thereto; andan auxiliary ultraviolet light emitting diode arranged at a site located at an outer periphery of the semiconductor wafer to apply an ultraviolet ray towards a peripheral edge of the protective tape. |
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052326560 | summary | BACKGROUND OF THE INVENTION This invention relates generally to the operation of fast acting nuclear reactor control devices and more particularly, to a device for providing primary drive to a safety control rod in a nuclear reactor to provide for control of the nuclear reaction, and as conditions warrant, for shutdown of the nuclear reactor. Nuclear reactors typically employ control rods which are inserted into the core of the reactor to control the level of the nuclear reaction. Control rods are commonly used during normal operation of the reactor to maintain a desired level of neutron flux in the core, and additionally, some of the safety rods provide a means for shutting down the reactor in emergency situations ("SCRAM"), or normal shutdown All such rods have a neutron absorbing portion containing a substance, such as hafnium, to control the flow of neutrons. Movement of the rods in or out of the core controls the nuclear reaction. When it is necessary to shut down the reactor during emergency situations the entire neutron absorbing portion of the safety control rod must be inserted as rapidly as possible into the reactor core. Various fast acting nuclear reactor control devices have been employed in attempting to provide a reactor control system. Most nuclear reactor facilities use spring drives held in a strained position. For example, one design employs a compression spring and a recirculating ball lead screw arrangement. The spring provides torque to the drive system mainshaft by driving the ball lead screw. This design has several limitations. The spring provides a decreasing torque resulting in less overall energy input and requires a high retaining torque. This encumbers the fast release capabilities of an electrical clutch also employed in this design, because high currents are required in the clutch to resist this torque. Another imitation is that the highly loaded lead screw is prone to galling and lacks efficiency. Additionally, the spring and ball screw SCRAM-assist system provides torque only during about the first one-third of the safety control rod's downward stroke as it is positioned in the reactor core. The remainder of the downward stroke is effected by the weight of the rod under the influence of gravity, and water pressure if so configured. Accordingly, it is an object of the present invention to overcome the inefficiencies of present nuclear reactor safety control rod drive units to improve the safety and reliability of nuclear reactor operation. It is a further object of the present invention to provide a fast acting nuclear reactor control device which improves the safety performance and reliability of nuclear reactor operation. Another object of the present invention is to provide a nuclear reactor control device which better maintains power and force to compel safety-rod insertion over the full length of the control rod. Yet another object of the present invention is to provide a nuclear reactor control device which provides for easy adjustment and control of the control rod. An additional object of the present invention is to provide a nuclear reactor control device which allows the safety control rod system to freely travel toward a safe position in the event of a partial drive system failure. SUMMARY OF THE INVENTION This invention provides a fast-acting nuclear reactor control device for controlling and positioning a safety control rod within the core of a nuclear reactor, the nuclear reactor being controlled by a reactor control system. The device includes a primary safety control rod drive means operatively connected to the safety control rod for driving and positioning the safety control rod within or without the reactor core. The safety control rod is oriented in a substantially vertical position to allow the safety control rod to fall into the reactor core under the influence of gravity and water pressure during shutdown of the reactor. The safety control rod is connected to a rack, and the primary drive means can be a safety control rod drive shaft having a pinion driving the rack for allowing the safety control rod to be positioned within or without the reactor core. The primary drive means is further operatively connected to a hydraulic pump such that operation of the primary drive means simultaneously drives the safety control rod to desired positions within the reactor core and operates the hydraulic pump such that a hydraulic fluid is forced into a pressurized accumulator, charging the accumulator with compressed gas for the storage of potential energy. A solenoid valve is interposed between the hydraulic pump and the accumulator, the solenoid valve being a normally open valve actuated to a closed position when the safety control rod is out of the reactor core during reactor operation. The solenoid valve opens in response to a signal from the reactor control system calling for shutdown of the reactor with rapid insertion of the safety control rod into the reactor core, such that the opening of the solenoid valve releases the potential energy in the accumulator to cause hydraulic fluid to flow back through the hydraulic pump, converting the hydraulic pump to a hydraulic motor having speed and power capable of full length insertion and high speed driving of the safety control rod into the reactor core. The insertion of the safety control rod will now be powered by the combined effects of high pressure gas, gravity, an differential water pressure. This will ensure rapid and reliable operation. The primary drive means or safety control rod drive shaft can include an electromagnetic clutch co-axial with the drive shaft for positioning the safety control rod in a run position. Further, the primary drive means includes an overrunning clutch co-axial with the drive shaft and located intermediate the hydraulic motor and the electromagnetic clutch. This overrunning clutch is capable of allowing the speed of the primary drive means to rotate at a speed greater than the speed of the hydraulic motor during shutdown of the reactor to provide for rapid insertion of the safety control rod into the reactor core. A reservoir of hydraulic fluid is connected to the hydraulic pump. The primary drive means further includes a drive motor driving through the electromagnetic clutch to position the safety control rod in a run position while simultaneously driving the hydraulic pump through the overrunning clutch. In this manner, hydraulic fluid is transferred from the reservoir to the accumulator. |
055090435 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS FIG. 1 shows an X-ray analysis apparatus with an X-ray source 1, a monochromator 3, a goniometer 5 and a detector 7 which are only diagrammatically shown. The X-ray source 1 comprises an anode 14 which is accommodated in a housing 10 provided with a radiation window 12, which anode consists of, for example copper, chromium, scandium or another customary anode material. An electron beam generates an X-ray beam 15 in the anode. The monochromator comprises two crystal pairs 18 and 20 with crystals 21, 23, 25 and 27. In the crystal pair 18 crystal end faces 22 and 24 serve as operative crystal faces. Similarly, in the crystal pair 20 crystal end faces 26 and 28 act as operative crystal faces. The first crystal pair can be arranged so as to be rotatable about an axis 30 extending perpendicularly to the plane of drawing, and the second crystal pair can be arranged similarly so as to be rotatable about an axis 32. The end faces 22, 24 and 26, 28 remain mutually parallel in any rotary position. Preferably, the crystals have, for each pair, a U-shape cut from a single monocrystal, the connecting portion of the U being used, for example for mounting the crystals. The inner faces of the limbs of the U then form the operative crystal end faces. Alter cutting and possibly grinding or polishing, a surface layer has been removed from these surfaces, for example by etching, in order to remove material in which stresses may have developed due to mechanical working. The carrier plate 34 for the monochromator has a comparatively rigid construction so that, for example its lower side can be used to support mechanical components, for example for the crystal orientation motions, without risking deformation of the plate. In the present embodiment, the length of one of the crystals of each of the crystal pairs is reduced so that more freedom is obtained in respect of a beam path. The attractive property of the 4-crystal monochromator as regards the angle of aperture for the incoming beam enables the X-ray source, i.e. actually a target spot on the anode 14, to be situated at a minimum distance from the first crystal pair, which minimum distance is determined by the construction of the source. An attractive intensity is thus achieved already for the ultimate analyzing X-ray beam 35. In the present embodiment the first crystal pair 18 is rotatable about the axis 30 of a shaft on which a first friction wheel 40 which is situated beneath the mounting plate is mounted so as to engage a second friction wheel 42 which is mounted on the shaft with the axis 32 about which the second crystal pair 20 is rotatable. However, the two crystal pairs may alternatively be mutually independently adjustable or the adjustment can be performed by means of a drive motor with, for example programmed settings adapted to the anode material to be used or to specimens to be analyzed. The crystals are preferably made of germanium having operative end faces which extend parallel to the (440) crystal faces of a germanium monocrystal which is relatively free from dislocations. By diffraction from the (440) crystal faces an extremely well monochromatized beam having, for example a relative wavelength width of 2.3.times.10.sup.-5, a divergence of, for example 5 arc seconds, and an intensity of up to, for example 3.times.10.sup.4 quants per second per cm.sup.2 can be formed. Such a sharply defined beam enables measurement of errors in lattice spacings of up to 1 to 10.sup.5 can be measured and high-precision absolute crystal lattice measurements can also be performed thereby. The monochromatization of the X-my beam is realized in the monochromator by the central two reflections, i.e. the reflections from the crystal faces 24 and 28. The two reflections from the end faces 22 and 26 do influence the beam parameters, but they guide the beam 35 in the desired direction coincident with the prolongation of the incoming beam 15. Wavelength adjustment is achieved by rotating the two crystal pairs in mutually opposite directions; during this motion, therefore, the position of the emergent beam 35 does not change. An intensity which is, for example 30 times higher can be achieved by utilizing reflections from (220) crystal faces, in which case a larger spread in wavelength and a larger divergence occur. The monochromator is non-rotatably connected to the goniometer 5 in which a specimen 46 to be analyzed is accommodated in a specimen holder 44. For the detection of radiation emerging from the specimen 46 there is provided a detector 7 which is rotatable along a goniometer circle 48 in known manner. The detector enables measurements to be made throughout a larger angular range and for different orientations of the specimen. For exact determination of the position and possible repositioning of the specimen, the goniometer may include an optical encoder which is not shown in the drawing. FIG. 2b shows an example of an asymmetrical system of crystals in accordance with the invention, compared with a similar symmetrical system as shown in FIG. 2a, comprising notably germanium crystals with (440) and (220) lattice planes, respectively. FIG. 2a shows the symmetrical system comprising crystals 21, 23, 25 and 27 in which the lattice planes extend parallel to crystal end faces 22, 24, 26 and 28, respectively. FIG. 2b shows an asymmetrical crystal system in which the lattice planes are chosen to extend parallel to the outwards facing end faces 40, 42, 44 and 46 of the crystals 23, 21, 27 and 25, respectively; however, the inwards facing crystal end faces 22, 24, 26 and 28 no longer extend parallel to the lattice planes in this Figure. Each crystal exhibits (220) as well as (440) lattice planes; in the upper crystal pairs of the FIGS. 2a and 2b the (440) lattice planes are used, whereas in the lower crystal pairs of the FIGS. 2a and 2b the (220) lattice planes are used. An incoming X-ray beam 15 emerges from the crystal system as a beam 35 which is collinear with the incident beam in all situations. A comparison of the beam diameter of the FIGS. 2a and 2b already demonstrates that the difference between the symmetrical and the non-symmetrical system is comparatively small for the (440) crystal planes, whereas it is substantial for the (220) crystal planes. The same holds for the resolution. |
abstract | A compact pressurised water nuclear reactor comprises a primary circuit fully integrated into the reactor vessel (10). Thus, a single steam generator (12) forms the cover of the vessel (10) and the pressuriser (30) and the primary pumps (28) are housed in the vessel (10). The same is true for the control mechanisms of the control rods (40). Finally, a venturi system (44) is also provided in the vessel (10) to create water circulation if there is a failure of the primary pumps (28). |
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abstract | An x-ray device for recording projection images of a patient features a C-arm on the ends of which an x-ray source and an x-ray detector are accommodated. The x-ray detector is accommodated on the C-arm so that translation movements can be executed with the x-ray detector in relation to the C-arm. This allows the angulation area or the area of the x-ray device with which an image can be recorded to be increased. |
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abstract | A system for simulating maintenance of a reactor core protection system that has at least two or more channels, includes: a simulation signal generation unit for generating a simulation state signal including a normal state or an abnormal state, a communication unit connected to each of the channels of the reactor core protection system to transmit the state signal to the channel, and a control unit for receiving a result signal output from the channel in response to the input simulation state signal and confirming whether the reactor core protection system normally determines a reactor core state by analyzing the result signal. |
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050842315 | claims | 1. A refueling mast for refueling a nuclear reactor, said mast comprising: plural longitudinally extending cylindrical tubes including an outermost tube and at least a first inner tube, said first tube having a first outer surface and being relatively dimensioned so that said inner tube can be slid within and relative to said outermost tube; said first inner tube having a first set of longitudinally extending tracks formed at said first outer surface, each of said tracks of said first set having longitudinally extending grooves; telescoping means for causing said first inner tube to slide within and relative to said outermost tube; and first guide means for guiding longitudinal movement of said first inner tube relative to said outermost tube, said first guide means including grooved rollers which mate with said first set of longitudinally extending tracks, said first guide means being mechanically coupled to said outermost tube so as to maintain a fixed longitudinal position relative thereto. a second inner tube, said second inner tube having a second outer surface and having a second set of longitudinally extending tracks formed at said second outer surface, each of said tracks of said second set having longitudinally extending grooves, each of said second set of tracks being radially aligned with a respective one of said set of tracks; and second guide means for guiding longitudinal movement of said second inner tube relative to said outermost tube, said second guide means including grooved rollers which mate with said second set of longitudinally extending tracks, said second guide means being mechanically coupled to an outer tube so as to maintain a fixed longitudinal position relative thereto. plural longitudinally extending cylindrical tubes including an outermost tube and plural inner tubes, said inner tubes including a first inner tube relatively dimensioned so that said first inner tube can be slid within and relative to said outermost tube, and a second inner tube relatively dimensioned so that said second inner tube can be slid within and relative to said first inner tube, said first inner tube having a first outer surface and said second inner tube having a second outer surface; said first inner tube having a first set of longitudinally extending tracks formed therein at said first outer surface; said second inner tube having a second set of longitudinally extending tracks formed therein at said second outer surface, each of said tracks of said first and second sets having longitudinally extending grooves, each of said second set of tracks being radially aligned with a respective one of said first set of tracks; hoist means for causing each inner tube to move within and relative the next radially outward and adjacent of said tubes; guide means for guiding longitudinal movement of each of said inner tubes, said guide means for each of said inner tubes including grooved rollers mounted on the next radially outward and adjacent of said tubes, each of said grooved rollers mating with a respective one of said grooved tracks; a fuel grapple attached to one of said inner tubes, said fuel grapple being adapted for engaging and lifting fuel elements; and transit means for moving said tubes laterally; whereby said hoist means can cause said inner tubes to protract and retract relative to said outermost tube so as to raise and lower said fuel grapple, said transit means providing for movement of fuel elements between a storage area and a reactor core in a fission reactor complex. 2. A refueling mast as recited in claim 1 further comprising: 3. A refueling mast as recited in claim 1 wherein said first inner tube is flattened in the location of said tracks. 4. A refueling system comprising: 5. The assembly of claim 4 wherein said inner tubes are flattened in the location of said tracks. |
059498390 | claims | 1. A fuel assembly for a boiling water reactor, wherein the fuel assembly during operation is arranged vertically in the core of the reactor and the fuel assembly comprises: a plurality of vertical fuel rods arranged in at least one fuel bundle, wherein each of the fuel rods comprises a stack of fuel pellets surrounded by a cladding tube, a bottom tie plate which retains and supports the lower part of the fuel bundle and is arranged at the bottom of the fuel assembly, a number of spacers which retain and position the fuel rods in spaced relationship to each other and are arranged axially separated along the fuel rods, a vertical water channel through which water flows upwardly through the fuel assembly, and a fuel channel which surrounds the fuel bundle, a top spacer which retains and supports the upper part of the fuel bundle and is arranged above and spaced apart from the stacks of fuel pellets, wherein 2. A fuel assembly according to claim 1, wherein the plenum tubes run through the top spacer and extend somewhat above the top spacer. 3. A fuel assembly according to claim 2, wherein the plenum tubes extend at least 5 cm above the top spacer. 4. A fuel assembly according to claim 1, wherein the upper part comprises a top piece provided with a handle, and the water channel terminates below and at a distance from the top piece. 5. A fuel assembly according to claim 4, wherein the water channel terminates above the top spacer. 6. A fuel assembly according to claim 4, wherein between the water channel and the top piece, an interconnecting member is arranged. 7. A fuel assembly according to claim 1, wherein the water channel is provided with inflow openings in its upper part. 8. A fuel assembly according to claim 1, wherein the water channel has a cruciform cross section. 9. A fuel assembly according to claim 1, wherein in the upper part of the plenum tube, a top plug is arranged which comprises a cylindrical pin having a diameter which is smaller than the diameter of the plenum tube. |
abstract | An apparatus directing x-rays along a predetermined axis includes an x-ray optic having one or more nested x-ray reflector rings positioned relative to a source generating broad spectrum x-rays so that generated x-rays moving away from the predetermined axis are collected by the reflector incident at or close to a Bragg angle to thereby reflect the collected x-rays into a conically parallel beam. A first diffractor is positioned relative to the x-ray optic to receive incident thereon the conically parallel beam, the first diffractor selected from a truncated cone and a cylinder and diffracting the conically parallel beam toward the predetermined axis. A second diffractor is positioned relative to the first diffractor and having a geometry effective to receive incident thereon and redirect the conically parallel beam along the predetermined axis as a collimated beam of substantially parallel x-rays. |
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abstract | System and method for XRD-based threat detection. An object is scanned with a first threat detection system. One or more alarm objects are identified. Data about the one or more alarm objects is passed from the first threat detection system to a second threat detection system and is used to move and/or to rotate the object in a predetermined ray path that decreases attenuation of scattered x-ray radiation. Also disclosed is a secondary collimator for XRD-based false alarm resolution in computed tomography {“CT”) threat detection systems. The secondary collimator comprises one or more slit apertures configured to provide a multi-angle capability that extends a range of momenta for which XRD intensities are measured for a predetermined range of photon intensities. |
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061809519 | claims | 1. A process of irradiation comprising the steps of: providing a beam of radiation having an energy and a direction of scan sweep; providing a reel having a center axis, the reel including a core substantially transparent to the beam of radiation; disposing around the reel a target material having a thickness; rotating the reel around the center axis; and directing the beam at the target material such that the direction of scan sweep is substantially perpendicular to the center axis, whereby the beam of radiation encounters the target material on a frontside of the reel, passes through the core, and reencounters target material on a backside of the reel such that the target material receives a substantially constant dose of radiation throughout its thickness. maintaining constant the speed of rotation, an energy of the radiation beam, a density of the target material, and a diameter of the core; and varying a thickness of the target material to produce a substantially constant dose of radiation throughout the thickness of the target material. maintaining constant the speed of rotation, an energy of the radiation beam, a density of the target material, and a thickness of the target material; and varying a diameter of the core to produce a substantially constant dose of radiation throughout the thickness of the target material. maintaining constant the speed of rotation, a diameter of the core, a density of the target material, and a thickness of the target material; and varying an energy of the radiation beam to produce a substantially constant dose of radiation throughout the thickness of the target material. maintaining constant a diameter of the core, a density of the target material, a thickness of the target material, and the energy of the radiation beam; and varying the speed of rotation to produce a substantially constant dose of radiation throughout the thickness of the target material. an radiation source producing a beam of radiation having a scan direction; a cylindrical reel having a core and a central axis, the core composed of material substantially transparent to the beam of radiation, the central axis substantially perpendicular to the scan direction, and the reel rotatable about the central axis; and a target material disposed around the cylindrical reel. 2. The process according to claim 1 wherein the substantially constant dose of radiation is such that the highest dose of radiation received by the target material is 10% or less of a dose of radiation received at a surface of the target material. 3. The process of irradiation according to claim 1 wherein the beam of radiation is x-ray radiation. 4. The process of irradiation according to claim 1 wherein the beam of radiation is gamma radiation. 5. The process of irradiation according to claim 1 wherein the beam of radiation is electron beam radiation. 6. A method of optimizing an irradiation process in which a target material is rotated at a speed on a core substantially transparent to a beam of radiation, the method comprising the steps of: 7. The method according to claim 6 wherein the substantially constant dose of radiation is such that the highest dose of radiation received by the target material is 10% or less of a dose of radiation received at a surface of the target material. 8. The method according to claim 6 wherein the beam of radiation is x-ray radiation. 9. The method according to claim 6 wherein the beam of radiation is gamma radiation. 10. The method according to claim 6 wherein the beam of radiation is an electron beam. 11. A method of optimizing an irradiation process in which a target material is rotated at a speed on a core substantially transparent to a beam of radiation, the method comprising the steps of: 12. The method according to claim 11 wherein the substantially constant dose of radiation is such that the highest dose of radiation received by the target material is 10% or less of a dose of radiation received at a surface of the target material. 13. The method according to claim 11 wherein the beam of radiation is x-ray radiation. 14. The method according to claim 11 wherein the beam of radiation is gamma radiation. 15. The method according to claim 11 wherein the beam of radiation is an electron beam. 16. A method of optimizing an irradiation process in which a target material is rotated at a speed on a core substantially transparent to a beam of radiation, the method comprising the steps of: 17. The method according to claim 16 wherein the substantially constant dose of radiation is such that the highest dose of radiation received by the target material is 10% or less of a dose of radiation received at a surface of the target material. 18. The method according to claim 16 wherein the beam of radiation is x-ray radiation. 19. The method according to claim 16 wherein the beam of radiation is gamma radiation. 20. The method according to claim 16 wherein the beam of radiation is an electron beam. 21. A method of optimizing an irradiation process in which a target material is rotated at a speed on a core substantially transparent to a beam of radiation, the method comprising the steps of: 22. The method according to claim 21 wherein the substantially constant dose of radiation is such that the highest dose of radiation received by the target material is 10% or less of a dose of radiation received at a surface of the target material. 23. The method according to claim 21 wherein the beam of radiation is x-ray radiation. 24. The method according to claim 21 wherein the beam of radiation is gamma radiation. 25. The method according to claim 21 wherein the beam of radiation is an electron beam. 26. An apparatus for irradiating a target material comprising: 27. The apparatus according to claim 26 wherein the source produces a beam of x-ray radiation. 28. The apparatus according to claim 26 wherein the source produces a beam of gamma radiation. 29. The apparatus according to claim 26 wherein the source produces an electron beam. |
abstract | A method of loading nuclear fuel assemblies into a fuel rack in an underwater (or other submerged) environment that reduces the depth required for the pool to effectuate the fuel rack loading procedure. In one embodiment, the method comprises submerging a nuclear fuel assembly having an axis and a horizontal cross-section in a pool; providing a fuel rack in the pool, the fuel rack comprising a body structure comprising at least one elongated cell, a top, a bottom, a first lateral side, at least one elongated slot in the first lateral side that forms a lateral passageway into the cell; positioning the fuel assembly laterally adjacent to the elongated slot of the fuel rack so that the axis of the fuel assembly is substantially aligned with the elongated slot; and translating the fuel assembly in a lateral direction through the elongated slot and into the cell. |
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043022910 | abstract | A structure for an underwater nuclear power generating plant comprising a triangular platform formed of tubular leg and truss members upon which are attached one or more large spherical pressure vessels and one or more small cylindrical auxiliary pressure vessels. |
045171521 | description | DETAILED DESCRIPTION OF THE INVENTION Numeral 1 refers to a fuel element bundle which comprises fuel elements 2 and guide tubes 3. Each fuel element 2 basically comprises a tubular jacket, the fuel tube, which is filled with radioactive material. Water fills the space between the fuel elements 2 and the guide tubes 3. For testing the fuel element tubes, ultrasonic transducers 4 and 5 disposed on respective finger shaped supports 6 and 7 are moved between the fuel elements 2 to cause the transmit transducer 4 to transmit ultrasonic waves into the tube to produce a revolving (circumferential) echo signal (position I), see zig-zag signal path in FIG. 3a. This echo signal is received by the receive transducer 5 and the resulting electrical signal is evaluated by the test instrument 8 and displayed on the screen of a cathode ray tube 9. Conductors 10 and 11 connect the transmit transducer and the receive transducer to the instrument 8. Numerals I to IV depict the different positions of the transducers as well as the corresponding echo indications on the cathode ray tube screen 9. In position I there can be seen the transmit pulse SI, the revolving echo signal UE, as well as the transmitted signal DE arising from the ultrasonic transmit pulse passing from the transmit transducer 4 directly to the receive transducer 5, see dashed signal path in FIG. 3a. When performing dynamic testing during which the transmit transducer 4 and receive transducer 5 are moved continuously past the fuel elements 2, the transmitted signal DE is disturbing as it appears in close proximity to the revolving echo signal UE. Particularly, when a defective tube is tested and the amplitude of the revolving echo diminishes due to ultrasonic energy scatter into the interior of the fuel element, it is readily possible to have an erroneous indication on account of the presence of the transmitted signal. For this reason a time gate circuit is used within which, as far as possible, only revolving echo signals are received. This gated interval 12 is shown by dashed lines. As stated heretofore, the use of a gate circuit with constant time axis setting leads to erroneous readings. Because of the change of the spacing between the transducers along their path between the fuel elements, see dashed path of the finger like supports 6 and 7, it is possible that the transmitted signals may also fall within the time gate (position III) or, alternatively, the gate is not properly adjusted and the revolving echo signal falls outside the time gate. In accordance with the invention the start of the time gate is shifted in correspondence with the motion of the transducers 4 and 5 along their path through the spacing of the fuel elements. To this end, as seen from FIG. 2a, the position of the transmitted signal and, hence, the transit time T.sub.L of the ultrasonic pulses between the transmit transducer 4 and the receive transducer 5, is determined in the gap between always the last tested tube and the next to be tested tube. The start of the gate interval is then given by subtracting a constant value C from the measured transit time value T.sub.L. The constant value C is somewhat greater than the constant width B of the gate 12 and is selected so that the transmitted signal DE received when measuring the transit time falls just outside the gate. With the gate parameters determined as described, see FIG. 3a, the next fuel element 2 is tested. Thereafter, the same procedure is repeated for determining the setting of the time gate. FIGS. 2b and 3b show the echo signal displays corresponding to the position of the transducers in FIGS. 2a and 3b. The gate 12 is shown by dashed lines. FIG. 4 shows a circuit arrangement for practicing the method described heretofore. A clock 13 periodically provides trigger signals to a pulse generator 14 for providing electrical transmit pulses to the transmit transducer 4. The electrical signals corresponding to received ultrasonic echo signals provided by the receive transducer 5 are fed to the amplifier 20, a display means 14, a gate circuit 15 and a transit time measurement means 16. When the transducers 4 and 5 are in the position indicated in FIG. 2a, i.e. between the fuel elements, a control means 17 provides a connection by switch S between the transit time measurement means 16 and a storage means 18 for storing the start or opening of the gate, i.e. the measured transit time value T.sub.L minus an adjustable constant value C. The constant value C can be set into the storage means, for example, by means of an encoding switch, not shown. The gate circuit 15 remains inhibited during the step of determining the gate setting. When the transducers have advanced into the position illustrated by FIG. 3a for testing the fuel element 2, the control means 17 connects the transit time measurement means 16 via switch S to the evaluation unit 19. Also the gate circuit 15 is operative to cause signals occurring within the gated time interval 12 to pass to the evaluating unit 19. The constant gate width B is adjusted by means of an encoding switch, not shown. The gate circuit 15 is normally an AND gate as known to those skilled in the art. The method described hereinbefore is not limited to the testing of fuel element tubes and other tubular arrays in a nuclear reactor, but is useful also for testing other closely spaced tubular articles, such as the tubes of a heat exchanger. |
claims | 1. A method for monitoring an irradiation planning, the method comprising:providing an irradiation planning data set that is created for irradiating a moving target volume;providing a motion signal that simulates a motion of the target volume;irradiating, using an irradiation system, a phantom with an ion particle beam using control parameters stored in the irradiation planning data set and the motion signal, the phantom being configured for detecting a dose distribution deposited in the phantom during or after the irradiation;ascertaining the dose distribution deposited in the phantom;calculating an expected dose distribution on the basis of parameters that are related to the control of the irradiation system during the irradiation; andcomparing the ascertained dose distribution deposited in the phantom with the calculated expected dose distribution,wherein the phantom has a motion pattern that differs from the motion signal, andwherein irradiating the phantom comprises gating, rescanning, tracking, or any combination thereof, the gating comprising activating and deactivating the irradiating using the motion signal, the rescanning comprising building up a total dose in the phantom with multiple successive applications of partial doses at a same site, the tracking comprising deflecting the ion particle beam as a function of the motion signal. 2. The method as defined by claim 1, wherein the phantom is a moving phantom. 3. The method as defined by claim 1, wherein the comparison is made between a nonhomogeneous pattern in the dose distribution deposited in the phantom and an expected nonhomogeneous pattern. 4. The method as defined by claim 1, wherein the parameters used for calculating the expected dose distribution are control parameters that are stored in the irradiation planning data set and include the motion signal. 5. The method as defined by claim 1, wherein the parameters used for calculating the expected dose distribution comprise data that characterize an actual property of a treatment beam during the irradiation. 6. The method as defined by claim 5, wherein the parameters used for calculating the expected dose distribution are used for putting the dose distribution into relation with an imaging data set that is the basis of the irradiation planning data set. 7. The method as defined by claim 5, wherein the parameters used for calculating the expected dose distribution comprise data that characterize a location of the treatment beam, an applied number of particles of the treatment beam during the irradiation, or a combination thereof. 8. The method as defined by claim 1, wherein the motion signal is a virtual motion signal generated internally in a computer unit. 9. The method as defined by claim 1, wherein the motion signal is a motion signal detected by a motion detection device. 10. The method as defined by claim 1, wherein irradiating the phantom comprises irradiating a 3D phantom. 11. The method as defined by claim 10, wherein irradiating the phantom comprises irradiating a plurality of distinguishable regions, the plurality of distinguishable regions comprising materials with a different penetration depth for a particle beam. 12. The method as defined by claim 11, wherein calculating the expected dose distribution comprises taking the constitution of the phantom into account. 13. The method as defined by claim 1, wherein the method is for monitoring the irradiation planning in a particle therapy system, in which with a treatment beam, the dose distribution is depositable in a target object. 14. The method as defined by claim 1, wherein the phantom is constituted anthropomorphically. 15. The method as defined by claim 1, wherein irradiating the phantom comprises gating, rescanning, and tracking. 16. An apparatus for monitoring an irradiation planning, the apparatus comprising:a first device, the first device configured for providing an irradiation planning data set that is optimized for irradiating a moving target volume with an ion particle beam generateable by an irradiation system;a second device, the second device configured for furnishing a motion signal that simulates a motion of the target volume;a phantom operable to detect a dose distribution deposited in the phantom during or after irradiation with the ion particle beam, wherein the phantom has a motion pattern that differs from the motion signal; anda computer device configured to:calculate an expected dose distribution on the basis of parameters that are related to the control of the irradiation system during the irradiation; andcompare an ascertained dose distribution deposited in the phantom with the calculated expected dose distribution,wherein the phantom is irradiated with the ion particle beam using gating, in which the irradiation controlled by the motion signal is activated and deactivated, rescanning, in which a total dose in the phantom is built up by multiple successive applications of partial doses at a same site, tracking, in which the ion particle beam is deflected as a function of the motion signal, or any combination thereof. 17. The apparatus defined as by claim 16, wherein the apparatus is for monitoring the irradiation planning in a particle therapy system, in which with a treatment beam, the dose distribution is depositable in a target object. 18. An irradiation system comprising:an apparatus comprising:a first device, the first device configured for providing an irradiation planning data set that is optimized for irradiating a moving target volume with an ion particle beam generateable by an irradiation system;a second device, the second device configured for furnishing a motion signal that simulates a motion of the target volume;a phantom operable to detect a dose distribution deposited in the phantom during or after irradiation with the ion particle beam, wherein the phantom has a motion pattern that differs from the motion signal; anda computer device configured to:calculate an expected dose distribution on the basis of parameters that are related to the control of the irradiation system during the irradiation; andcompare an ascertained dose distribution deposited in the phantom with the calculated expected dose distribution,wherein the comparison is made between a nonhomogeneous pattern in the dose distribution deposited in the phantom and an expected nonhomogeneous pattern, andwherein the phantom is irradiated with the ion particle beam using gating, in which the irradiation controlled by the motion signal is activated and deactivated, rescanning, in which a total dose in the phantom is built up by multiple successive applications of partial doses at a same site, tracking, in which the ion particle beam is deflected as a function of the motion signal, or any combination thereof. 19. The irradiation system as defined by claim 18, further comprising a particle therapy system. |
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summary | ||
claims | 1. A turret-less x-ray monochromator apparatus, comprising:one or more x-ray reflectors that are configured to be selectively positioned in situ in sufficient proximity to a known location of a source of x-rays to collect the x-rays over a solid collection angle greater than 0.010 steradians, wherein the one or more reflectors are not mounted in a turret;a detector positioned to detect x-rays from the source that are reflected by the one or more reflectors; andmeans for scanning the one or more reflectors to reflect x-rays of different energies to the detector. 2. The apparatus of claim 1, further comprising a cassette configured to store one or more of the x-ray reflectors; a positioning mechanism located proximate the known location, the positioning mechanism being adapted to receive a reflector; and a transfer mechanism adapted to transfer a reflector from the cassette to the positioning mechanism and/or transfer a reflector from the positioning mechanism to the cassette. 3. The apparatus of claim 2 wherein the cassette is positioned at a sufficient distance from the positioning mechanism that the cassette does not interfere with an ability of the positioning mechanism to place a reflector close enough to the known location that the reflector can collect x-rays from the known location over a solid angle of about 0.010 steradians or greater. 4. The apparatus of claim 2 wherein the cassette and reflector positioner are housed in a vacuum chamber containing an aperture for the entry of x-rays from the known location. 5. The apparatus of claim 4 wherein the vacuum chamber is attached to a source of the x-rays originating at the known location. 6. The apparatus of claim 5 wherein the source of x-rays is an electron probe micro-analysis (EPMA) system having an electron beam column that focuses x-rays onto a surface of a sample, wherein the known location is on the surface of the sample. 7. The apparatus of claim 6 wherein the vacuum chamber is attached to the EPMA system so that the entry aperture is aligned with an x-ray detector port of the EPMA system. 8. The apparatus of claim 2 wherein one or more x-ray reflectors include one or more scannable circular cylindrical x-ray reflectors. 9. The apparatus of claim 2 wherein the one or more x-ray reflectors include one or more non-scannable reflectors. 10. The apparatus of claim 9 wherein the one or more non-scannable reflectors are optimized for detection of individual elements. 11. The apparatus of claim 9 wherein the non-scannable detectors have one or more of the following properties: multilayer coatings that are not efficient at more than one narrow energy range, multilayer coatings that have a graded interlayer spacing (d-spacing) for increased solid angle collection, elliptical cylindrical shapes, and doubly curved ellipsoidal shapes. 12. The apparatus of claim 2 wherein the x-ray reflectors could include more than 5 reflectors. 13. The apparatus of claim 2 wherein the positioning mechanism receives a reflector from the cassette and places the reflector into position for reflecting x-rays from the known location. 14. The apparatus of claim 2 wherein the positioning mechanism is adapted to scan the reflector during a spectrum measurement. 15. The apparatus of claim 14 further comprising a base plate attached to the positioning mechanism, wherein a mask containing an aperture and a detector are mounted to the base plate. 16. The apparatus of claim 15 wherein the mask is removable mounted to the base plate, whereby masks having different sized apertures can be mounted to the base plate. 17. The apparatus of claim 16 wherein the aperture is slightly larger than a focal spot size produced by the reflector and of the same shape as the spot size produced by the reflector. 18. The apparatus of claim 16 further comprising an aperture selector mechanism mounted to the detector, the aperture selector mechanism having two or more different apertures that can be selectively interposed between the reflector and the detector. 19. The apparatus of claim 2, further comprising a cassette positioner, wherein the cassette positioner is configured to move the cassette to bring a reflector location in the cassette in line with the transfer mechanism. 20. The apparatus of claim 17 wherein the transfer mechanism is configured to transfer a reflector from the reflector location to the positioning mechanism and/or vice versa by linear motion of the reflector. 21. The apparatus of claim 2 wherein the cassette includes locations for storing five (5) or more reflectors. 22. The apparatus of claim 21 wherein the five or more reflectors include between about seven (7) and about twenty (20) reflectors. 23. The apparatus of claim 21 wherein the five or more reflectors include different reflectors with a different reflector for every K, L and M x-ray line between about 0.15 keV and about 2.0 keV. 24. An x-ray monochromator apparatus, comprising:a non-focusing x-ray reflector characterized by a shape that does not focus x-rays, the non-focusing x-ray reflector being positioned in sufficient proximity to a known location of a source of x-rays to collect the x-rays over a solid collection angle and reflect the x-rays towards a detector. 25. The apparatus of claim 24, wherein the detector is a large area detector configured to intercept x-rays from the known location that are reflected by the non-focusing reflector. 26. The apparatus of claim 25 wherein the detector is adapted to determine a position where an x-ray lands on the detector, whereby the detector is a position sensitive detector. 27. The apparatus of claim 26 wherein the position sensitive detector is a semiconductor array, charge-coupled-device (CCD), multiwire proportional counter or other position sensitive x-ray detector. 28. The apparatus of claim 25 wherein the detector is a gas-filled proportional counter or a solid state detector. 29. The apparatus of claim 25 wherein the large area detector is in the form of an array of detectors. 30. The apparatus of claim 24 wherein the non-focusing reflector is an x-ray optic in the shape of a circular cylinder with the known location being located on a central axis of the cylinder. 31. The apparatus of claim 30 wherein the cylinder is in a fixed position with respect to the known location and detector. 32. The apparatus of claim 30 wherein an inside surface of the optic has a multilayer coating. 33. The apparatus of claim 32 wherein a d-spacing of the coating is graded to maintain a correct Bragg angle along the length of the optic. 34. The apparatus of claim 30, further comprising an aperture stop configured to prevent unreflected x-rays and stray electrons from entering the detector. 35. The apparatus of claim 30 wherein a maximum length of the optic is 2w, where w is a working distance between a front end of the optic and the known location of the source of x-rays. 36. The apparatus of claim 30 wherein the cylinder is manufactured in two or more longitudinal sections. 37. The apparatus of claim 30 wherein the cylinder is configured to perform a radial scan by changing a radius of the cylinder. 38. The apparatus of claim 37, further comprising an aperture stop configured to prevent unreflected x-rays and stray electrons from entering the detector wherein the aperture stop is configured to move axially toward the known location as the cylinder radius is increased in order to prevent unreflected x-rays from passing through the new enlarged cylinder. 39. The apparatus of claim 37 wherein the cylinder is made of multiple cylindrical segments configured to move radially with respect to a central axis of the cylinder. 40. The apparatus of claim 39 wherein each cylindrical segment is rigid and does not change its curvature as it moves radially with respect to the central axis, and the resulting reflecting surface of the expanded array does not lie on a circle. 41. The apparatus of claim 39 wherein the multiple cylindrical segments include four or more segments. 42. The apparatus of claim 39 wherein the multiple cylindrical segments include six or more segments. 43. The apparatus of claim 24 wherein the non-focusing optic includes a conical-shaped optic, wherein the conical shaped optic is characterized by an axis of conical symmetry. 44. The apparatus of claim 43 wherein the known location is located on the axis of conical symmetry. 45. The apparatus of claim 24 wherein the non-focusing optic has horn-like shape that is symmetric about a central axis. 46. The apparatus of claim 45 wherein the horn-like shape is selected such that an axial scan of the non-focusing optic allows different energy x-rays from the known location to be reflected to the detector. 47. An electron probe micro-analysis (EPMA) system, comprising:an electron beam column configured to focus a beam of electrons at a known location on a surface of a sample;an x-ray detector; and a non-focusing x-ray reflector disposed between the surface of the sample and the x-ray detector, the non-focusing x-ray reflector being configured to reflect x-rays originating at the sample surface toward the detector, wherein the non-focusing x-ray reflector is characterized by a shape that does not focus the x-rays. 48. An electron probe micro-analysis (EPMA) system, comprising:an electron beam column configured to focus a beam of electrons at a known location on a surface of a sample;an x-ray detector;a cassette configured to store one or more of the x-ray reflectors; a positioning mechanism disposed between the cassette and the sample, the positioning mechanism being configured to receive an x-ray reflector and position the reflector proximate the known location; and a transfer mechanism adapted to transfer a reflector from the cassette to the positioning mechanism and/or transfer a reflector from the positioning mechanism to the cassette. |
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051046113 | claims | 1. In a heat exchanger having a vessel containing a tubesheet which separates a primary side cavity and a secondary side cavity, tubes mounted in the tubesheet in fluid communication with the primary side cavity and extending through the tubesheet and into the secondary side cavity to provide a heat exchanging relationship between fluid in the tubes and fluid in the secondary side cavity, said secondary side cavity including a flow divider plate an having edges attached to the wall of the vessel and extending toward the tubesheet perpendicularly therewith, and a handhole through the vessel adjacent the divider plate perpendicular edges for maintenance and cleaning in the secondary side cavity and a closure plate for the handhole when the heat exchanger is in operation, the improvement comprising: said divider plate having a notch in said edge, said handhole being in alignment with said notch, a flow blocker extending from said handhole into said notch dimensioned such that it substantially fills said handhole and said notch and thereby with the flow divider plate prevents substantial flow from one side of the flow divider plate to the other when the heat exchanger is in operation and such that it can be removed from said vessel through said handhole to permit access to both sides of said divider plate for maintenance and cleaning. an inner cylindrical member with a recessed end portion, a hollow outer cylindrical member of greater diameter than thickness X into which said inner member is telescopically assembled, a spring biasing said inner member and said outer member in opposite directions, a longitudinal member limiting the longitudinal telescoping extension of the two members, an end portion of said longitudinal member being located in said recessed end portion of said inner cylindrical member, said recessed end portion of said inner cylinder being of an overall transverse dimension not substantially greater than X. 2. The improved heat exchanger of claim 1 in which the handhole is adjacent the tubesheet in the secondary side cavity and the divider plate intersects the tubesheet. 3. The improved heat exchanger of claim 1 in which the handhole and the flow blocker are substantially cylindrical and the opposed notch edges adjacent the vessel are spaced by the diameter of the cylindrical flow blocker. 4. The improved heat exchanger of claim 3 in which the opposed notch edges adjacent the vessel are arcuate and concentric with the cylindrical flow blocker and the handhole. 5. The improved heat exchanger of claim 1 in which the flow blocker is attached to the closure plate for the handhole. 6. The improved heat exchanger of claim 1 in which the flow blocker is longitudinally compressible to maintain it in position between the notch and the closure plate. 7. The improved heat exchanger of claim 6 in which the flow blocker comprises two substantially cylindrical members, an inner member and an outer member telescopically assembled and biased in opposite directions by spring means. 8. The improved heat exchanger of claim 7 in which the flow blocker has a longitudinal member limiting the longitudinal telescopic extension of the two members. 9. The improved heat exchanger of claim 1 in which the flow blocker is attached to the closure plate for the handhole. 10. The improved heat exchanger of claim 1 in which the divider plate is part of an economizer of a nuclear steam generator. 11. A flow blocker for closing a passage in a nuclear steam generator having a handhole and a secondary side divider plate of a thickness X comprising in combination: 12. The flow blocker of claim 11 which includes means attaching the outer cylindrical member to a closure plate for a handhole of the nuclear steam generator. |
description | This is a continuation-in-part of co-pending U.S. patent application Ser. No. 10/383,096, filed on Mar. 5, 2003, which is a continuation-in-part of co-pending U.S. patent application Ser. No. 09/932,531, filed on Aug. 17, 2001, now U.S. Pat. No. 7,231,011 both of which are hereby incorporated herein by reference for all that they disclose. The United States Government has rights in the following invention pursuant to Contract No. DE-AC07-99ID13727 between the U.S. Department of Energy and Bechtel BWXT Idaho, LLC. This invention relates generally to the testing and evaluation of materials and more specifically to methods and apparatus for performing non-destructive testing of materials using position annihilation. Non-destructive material evaluation refers to any of a wide variety of techniques that may be utilized to examine materials for defects and/or evaluate the materials without requiring that the materials first be destroyed. Such non-destructive material evaluation is advantageous in that all materials or products may be tested for defects. After being evaluated, acceptable (e.g., substantially defect-free or with acceptable defect levels) materials may be placed in service, while the defective materials may be re-worked or scrapped, as may be required. Non-destructive evaluation techniques are also advantageous in that materials already in service may be evaluated or examined in-situ, thereby allowing for the early identification of materials or components that may be subject to in-service failure. The ability to evaluate or examine new or in-service materials has made non-destructive material evaluation techniques of great importance in safety- or failure-sensitive technologies, such as, for example, in conventional aviation and space technologies, as well as in nuclear systems and in power generation systems. One type of non-destructive evaluation technique, generally referred to as positron annihilation, is particularly promising in that it is theoretically capable of detecting fatigue and other types of damage in metals at its earliest stages. While several different positron annihilation techniques exist, as will be described below, all involve the detection of positron annihilation events in order to ascertain certain information about the material or object being tested. By way of background, complete annihilation of a positron and an electron occurs when both particles collide and their combined mass is converted into energy in the form of two (and occasionally three) photons (e.g., gamma rays). If the positron and the electron are both at rest at the time of annihilation, the two gamma rays are emitted in exactly opposite directions (e.g., 180° apart) in order to satisfy the requirement that momentum be conserved. Each annihilation gamma ray has an energy of about 511 keV, the rest energies of an electron and a positron. In positron annihilation analysis, the momentum of the positron is related to the environment in which it resides. For example, positron momentum is relatively low in defects (e.g., microcracks in composite materials and polymers) or in large lattice structures, whereas positron momentum is higher in defect-free or tight lattice structures. One way to determine the momentum of the positron is to measure the degree of broadening of the gamma energy line caused by the annihilation event. Alternatively, the momentum of the positron may be derived from the deviation from 180° of the annihilation gamma rays. Additional information about the electron density of the material at the site of annihilation may be obtained by determining the average lifetime of the positrons before they are annihilated. Still other information about the annihilation event may be detected and used to derive additional or supplemental information regarding the material being tested, such as the presence of contaminants or pores. Accordingly, the detection of positrons and the products of annihilation events provide much information relating to defects and other characteristics of the material or object being tested. As mentioned above, several different positron annihilation techniques have been developed. In one type of positron annihilation technique, positrons from a radioactive source (e.g., 22Na, 68Ge, or 58Co) are directed toward the material to be tested. Upon reaching the material, the positrons are rapidly slowed or “thermalized.” That is, the positrons rapidly loose most of their kinetic energy by collisions with ions and free electrons present at or near the surface of the material. After being thermalized, the positrons then annihilate with electrons in the material. During the diffusion process, the positrons are repelled by positively-charged nuclei, thus tend to migrate toward defects such as dislocations in the lattice sites where the distances to positively-charged nuclei are greater. In principle, positrons may be trapped at any type of lattice defect having an attractive electronic potential. Most such lattice defects are so-called “open-volume” defects and include, without limitation, vacancies, vacancy clusters, vacancy-impurity complexes, dislocations, grain boundaries, voids, and interfaces. In composite materials or polymers, such open-volume defects may be pores or microcracks. Generally speaking, positron annihilation techniques utilizing external positron sources are of limited utility in that the positrons from the external positron sources cannot penetrate very deeply into the materials. As a result, such techniques are limited to evaluating the surface structures of the materials being tested. Another type of positron annihilation technique replaces the external positron source with an external neutron source. Neutrons from the neutron source are directed toward the material being tested. Given sufficient energies, the neutrons will, in certain materials, result in the formation of isotopes that produce positrons. Such isotopes are commonly referred to as positron emitters, and include certain isotopes of copper, cobalt, and zinc. The positrons produced within the materials by the positron emitters then migrate to lattice defect sites, ultimately annihilating with electrons to produce gamma rays. This type of positron annihilation technique is often referred to as “neutron-activated positron annihilation” because it utilizes neutrons to trigger or induce the production of positrons. Neutron-activated positron annihilation techniques are advantageous over techniques that utilize external positron sources because the neutrons from the external neutron sources penetrate more deeply into the materials being tested than do positrons alone (e.g., from the external positron sources). Therefore, neutron-activated positron annihilation systems are generally capable of detecting flaws deep within the material rather than merely on the surface. Disadvantageously, however, neutron-activated positron annihilation techniques are limited to use with materials that contain positron emitters (i.e., certain isotopes of copper, cobalt, and zinc). A method for evaluating a material specimen comprises: Mounting a neutron source and a detector adjacent the material specimen; bombarding the material specimen with neutrons from the neutron source to create prompt gamma rays within the material specimen, some of the prompt gamma rays being emitted from the material specimen, some of the prompt gamma rays resulting in the formation of positrons within the material specimen by pair production; collecting positron annihilation data by detecting with the detector at least one emitted annihilation gamma ray resulting from the annihilation of a positron; storing the positron annihilation data on a data storage system for later retrieval and processing; and continuing to collect and store positron annihilation data, the continued collected and stored positron annihilation data being indicative of an accumulation of lattice damage over time. One embodiment of apparatus 10 for evaluating a material specimen 12 is illustrated in FIG. 1 and may comprise a neutron source or generator 14 and a detector assembly 16. The neutron source or generator 14 produces neutrons n and dire cts the neutrons n toward the material specimen 12. The neutrons n interact with the material specimen 12, resulting in the production of prompt gamma rays γp. While some of the prompt gamma rays γp are emitted from the material specimen 12, others of the prompt gamma rays γp will result in the formation of positrons e+ within the material specimen 12 through a process known as “pair production,” (illustrated schematically in FIG. 1 at 18) More specifically, and as will be described in greater detail herein, prompt gamma rays γp having energies greater than about 1.1 MeV are very likely to produce positrons e+ within the material specimen 12. Many of the positrons e+ produced as a result of the pair production process ultimately annihilate with electrons e− within the material specimen 12. The annihilation event results in the formation of annihilation gamma rays γa. As mentioned above, some of the prompt gamma rays γp resulting from the neutron bombardment of the material specimen 12 are emitted from the material specimen 12 and are detected by the detector assembly 16. In addition, some of the annihilation gamma rays γa formed as a result of the annihilation of positrons e+ and electrons e− are emitted from the material specimen 12 and are also detected by the detector assembly 16. The detector assembly 16 produces prompt gamma ray data 20 based on the detected prompt gamma rays γp and positron annihilation data 22 based on the detected annihilation gamma rays γa. A data processing system 24 operatively associated with the detector assembly 16 processes the prompt gamma ray data 20 and the positron annihilation data 22 in accordance with certain algorithms (described below) in order to produce output data that are indicative of a lattice characteristic of the material specimen 12. For example, in one embodiment, the data processing system 24 processes the prompt gamma ray data 20 and the positron annihilation data 22 in accordance with a positron lifetime algorithm 38 (FIG. 2) to produce positron lifetime data. Because the density of electrons is lower in defects contained in a material specimen compared to a defect-free material specimen, the mean lifetime of positrons trapped in defects is longer than those contained in a defect-free material. Further additional information on the sizes of the defects and other information, such as oxide inclusions, lattice structure variations, or localized composition changes, can be derived from the positron lifetime. Therefore, the positron lifetime data will be indicative of the presence of certain defects in the material specimen 12. Thereafter, the positron lifetime data and/or information relating to the presence of defects in the material specimen 12 may be presented in human-readable form on a suitable display system 26. The data processing system 24 may also be provided with a Doppler-broadening algorithm 40 (FIG. 2). The Doppler-broadening algorithm 40 is used to determine the degree of broadening of the gamma energy line (i.e., the 511 keV peak) of the detected annihilation gamma rays γa. The degree of broadening of the 511 keV peak is related to the momentum of the positron involved in the annihilation event. Therefore, the Doppler-broadening algorithm 40 may be used to assess certain characteristics associated with lattice defects contained in the material specimen 12, such as, for example, damage resulting from mechanical and thermal fatigue, embrittlement, annealing, or manufacturing defects. The resulting output data from the Doppler-broadening algorithm and/or information relating to certain lattice defects in the material specimen 12 may also be presented on the display system 26. A significant advantage of the present invention relates to the ability to produce the positrons within the bulk of the material specimen itself, rather than externally and to produce positron lifetime or Doppler broadening spectral data using the prompt gamma ray produced from the interaction of neutrons within the material. As a result, the method and apparatus of the present invention may be used to evaluate the lattice characteristics contained within the bulk of the material specimen, rather than merely on the surface. Another advantage of the method and apparatus of the present invention is that it has increased sensitivity over conventional positron annihilation techniques that utilize external positron sources in that there is little extraneous background “noise” caused by annihilations external to the specimen being analyzed. The increased sensitivity also allows other types of detectors (e.g., germanium, BaF2, or plastic) to be used. Moreover, the surface of the material specimen need not be specially prepared as is typically required with techniques that utilize external positron sources. Still yet another advantage of the invention is that it may be used with any of a wide range of material specimens, as positron formation by the process of pair production does not require the material specimen to contain positron emitters, as is required if the positrons are to be formed via the process of neutron activation. Consequently, the present invention may be used in conjunction with a practically unlimited variety of material specimens. Having briefly described one embodiment 10 of apparatus for evaluating a material specimen, as well as some of its more significant features and advantages, the various embodiments of methods and apparatus for evaluating a material specimen according to the present invention will now be described in detail. With reference now specifically to FIG. 1, one embodiment of apparatus 10 for evaluating a material specimen 12 may comprise a neutron source 14 for directing neutrons n toward the material specimen 12. As discussed above, the neutrons n from the neutron source 14 interact with the material specimen 12 and result in the production of prompt gamma rays γp within the material specimen 12. While some of the prompt gamma rays γp are emitted from the material specimen 12, others of the prompt gamma rays γp will result in the formation of positrons e+ within the material specimen 12 through the process of pair production. Many of the positrons e+ produced as a result of the pair production process ultimately annihilate with electrons e− within the material specimen 12. The annihilation event results in the formation of annihilation gamma rays γa, most of which are thereafter emitted from the material specimen 12. Before proceeding, it should be noted that, in addition to the formation of positrons e+ via the process of pair production, described above, positrons e+ may also be formed within the material specimen 12 by a process known as “neutron activation” if the material specimen 12 contains a positron emitter (not shown) capable producing positrons e+ in response to neutron bombardment. However, positron formation via the process of neutron activation is not of primary importance in the present invention and is not a significant component of the measurement response. In accordance with the teachings contained herein, it is generally preferred that the neutrons n from the neutron source 14 have energies in the range of about 0.1 MeV to about 4 MeV. In accordance with this requirement, any of a wide range of neutron sources, such as neutron generators or isotopic neutron sources, may be used in conjunction with the present invention. Examples of neutron generators include, but are not limited to, deuterium-deuterium (D-D) and deuterium-tritium (D-T) generators of the type that are well-known in the art and readily commercially available. An example of an isotopic neutron source includes, but is not limited to, 252Cf. In the embodiment shown in FIG. 1, the neutron source 14 comprises an isotopic neutron source 54, such as, for example 252Cf. The isotopic neutron source 54 may be surrounded by suitable shield 56 and reflector 58 to reduce stray neutron emission and to help direct additional neutrons n toward the material specimen 12. The shield 56 and reflector 58 may comprise any of a wide range of materials well-known in the art or that may be developed in the future that are or would be suitable for such uses, as would be obvious to persons having ordinary skill in the art after having become familiar with the teachings of the present invention. Consequently, the present invention should not be regarded as limited to a shield 56 and reflector 58 comprising any particular materials. However, by way of example, in one preferred embodiment, the shield 56 comprises lead, whereas the reflector 58 comprises carbon. In another embodiment, an additional reflector (not shown) may be positioned behind the material specimen 12 to direct neutrons n back toward the material specimen 12. It is generally preferred, but not required, to provide a moderator or thermalizer 60 between the neutron source 14 and the material specimen 12. The thermalizer 60 thermalizes the neutrons n from the neutron source 14, reducing their energies, thereby improving the number of interactions within the material specimen 12. Accordingly, the amount of thermilization to be provided will depend on the energies of the neutrons n from the neutron source 14, as well as on certain characteristics (e.g., thickness, density, etc.) of the material specimen 12, being studied. Generally speaking, it is preferred that the prompt gamma rays γp have energies of at least about 1.1 MeV, and preferably about 2.0 MeV, in order to produce high positron yields through the process of pair production. Because the energies of the prompt gamma rays γp are not related to the energies of the bombarding neutrons n, variations in the neutron energies will result in the number of neutrons deposited in the material specimen 12 which must be controlled depending on the type of material specimen 12 being examined. Therefore, the thermalizer 60 should be configured to allow the specimen 12 to be bombarded with neutrons having the appropriate energies for the material specimen 12 and thickness being examined. In one preferred embodiment, the thermalizer 60 comprises a material having a low atomic number, such as polyethylene. The overall length 62 of the polyethylene thermalizer 60 may be changed or varied as necessary to provide the desired degree of thermalization in accordance with the teachings provided herein. Alternatively, other types of thermalizers comprising other types of materials may be used, as would be obvious to persons having ordinary skill in the art after having become familiar with the teachings of the present invention. It is generally preferred, but not required, to provide additional shielding 64 around the thermalizer 60 in order to further reduce the amount of radiation from the neutron source 14 that may reach the detector assembly 16. The presence of such additional shielding 64 will enhance the sensitivity of the detector assembly 16 by reducing the amount of “background” radiation or noise detected by the detector assembly 16. By way of example, in one preferred embodiment, such additional shielding 64 may comprise any of a wide range of bismuth, lead, or borated polymer materials. The neutron source 14 is positioned adjacent the material specimen 12 to be tested so that neutrons n from the neutron source 14 are directed toward and bombard (i.e., penetrate) the portion of the material specimen 12 that is to be evaluated in accordance with the teachings of the present invention. In this regard it should be noted that any of a wide range of techniques may be used to irradiate the material specimen 12 with the neutrons n from the neutron source 14 so that the desired portions of the material specimen 12 are exposed to sufficient neutron flux to produce prompt gamma rays γp having sufficient energies to produce a high flux of positrons e+ through the process of pair production. Consequently, the present invention should not be regarded as limited to any particular technique for irradiating the material specimen 12. However, by way of example, in one preferred embodiment, the material specimen 12 may be irradiated by moving the specimen 12 and neutron source 14 with respect to one another so that the desired region on the material specimen 12 is exposed to neutron flux from the neutron source in amounts sufficient to produce prompt gamma rays γp having the desired energies, e.g., at least about 1.1 MeV and preferably about 2.0 MeV. The detector assembly 16 may be positioned adjacent the material specimen 12 so that the detector assembly 16 receives both prompt gamma rays γp and annihilation gamma rays γa emitted from the specimen 12. In one embodiment, the detector assembly 16 comprises a first detector 30 and a second detector 32 positioned in generally opposed, spaced-apart relation in the manner illustrated in FIG. 1. As will be described in greater detail below, the detectors 30 and 32 comprising the detector assembly 16 may be used to detect prompt gamma rays γp and/or annihilation gamma rays γa, depending on the particular algorithm (e.g., either the positron lifetime algorithm 38 or the Doppler broadening algorithm 40) that is being used to process the data. Therefore, it should be understood that the first detector 30 may produce prompt gamma ray data 20, positron annihilation data 22, or some combination of the two (if both prompt gamma rays γp and annihilation gamma rays γa are detected). Similarly, the second detector 32 may produce prompt gamma ray data 20, positron annihilation data 22, or some combination of the two. The first detector 30 may be provided with a collimator 34, such as a variable slit or other type of collimator, to collimate the gamma rays (e.g., the prompt gamma rays γp and/or the annihilation gamma rays γa, as the case may be) emitted by the material specimen 12. Similarly, the second detector 32 may be provided with a collimator 36 to collimate the gamma rays emitted by the material specimen 12. The collimator 36 may also comprise a variable slit collimator, although other types may be used. It should be noted that the detectors 30 and 32 comprising the detector assembly 16 need not be positioned in opposed, spaced-apart relation in the manner schematically illustrated in FIG. 1. Instead, the detectors 30 and 32 may be located with respect to the material specimen 12 in any of a wide range of positions, as may be necessary or desirable in any particular circumstance and as would be obvious to persons having ordinary skill in the art after becoming familiar with the teachings provided herein. Each detector 30 and 32 may comprise any of a wide range of detectors that are now known in the art or that may be developed in the future that are or would be suitable for detecting the prompt gamma rays γp and the annihilation gamma rays γa. Consequently, the present invention should not be regarded as limited to any particular type of gamma ray detector. However, by way of example, in one preferred embodiment, each detector 30 and 32 may comprise a germanium detector of the type that is well-known in the art and readily commercially available. Alternatively, other types of detectors, such as BaF2 or plastic-type detectors may be used. The data processing system 24 is operatively associated with the detector system 16 and receives the prompt gamma ray data 20 and positron annihilation data 22 produced by the detector system 16. As was briefly described above, the data processing system 24 processes the prompt gamma ray data 20 and positron annihilation data 22 in accordance with a positron lifetime algorithm 38. See FIG. 2. So processing the prompt gamma ray data 20 and the positron annihilation data 22 results in positron lifetime data. In addition, the data processing system 24 may also process the positron annihilation data 22 in accordance with the Doppler-broadening algorithm 40. The positron lifetime algorithm 38 is used to derive information regarding the characteristics of lattice defects contained in the material specimen 12. For example, the positron lifetime algorithm 38 may be used to obtain information as to whether the lattice defects comprise monovacancies, dislocations, slip zones, or particulate inclusions. In addition, information obtained from the mean lifetime of various defect components may be used to derive information relating to changing characteristics of the defects present in the specimen. The positron lifetime algorithm 38 basically involves a determination of an elapsed time between positron formation and positron annihilation. In order to do so, the positron lifetime algorithm utilizes the prompt gamma ray data 20 as well as the positron annihilation data 22. Because these data 20 and 22 are related to the prompt gamma ray γp associated with the formation of the positron e+, as well as the annihilation gamma rays γa produced by the positron annihilation event, respectively, the time between these two events is the positron lifetime. With reference now primarily to FIG. 3, the positron lifetime algorithm 38 may involve the use of both of the detectors 30 and 32 comprising the detector assembly 16 in order to determine positron lifetime. For example, in one operational sequence 66, the data processing system 24 monitors one of the detectors (e.g, detector 30) for prompt gamma ray data 20 at step 68. Upon detecting a prompt gamma ray γp, the data processing system 24 then monitors the other of the detectors (e.g., detector 32) and collects positron annihilation data 22 at step 70. The positron annihilation data 22 captured for a collection period that is between about 1 nanosecond (ns) to about 20 ns (12 ns preferred) after the detection of a prompt gamma ray γp (step 68). Positron annihilation data 22 collected during the collection period corresponds to annihilation events resulting from the same events that caused the production of the prompt gamma ray. The data processing system 24 then processes the prompt gamma ray data and positron annihilation data in order to determine positron lifetime at step 72. The operational sequence 66 may be achieved by providing the data processing system 24 with certain systems and devices illustrated in FIG. 4. More specifically, the data processing system 24 may be provided with a first timing discriminator 42 and a second timing discriminator 44. The first timing discriminator 42 is operatively connected to the first detector 30 of the detector assembly 16 and receives the prompt gamma ray data 20 produced by the first detector 30. The second timing discriminator 44 is operatively connected to the second detector 32 of the detector assembly 16 and receives positron annihilation data 22 from the second detector 32. The output 46 and 48 of each respective timing discriminator 42 and 44 is connected to a fast coincidence processor 50 and a time-to-amplitude converter 52 in the manner illustrated in FIG. 4. The combination of the timing discriminators 42 and 44, the fast coincidence processor 50, and the time-to-amplitude converter 52 allow the data processing system 24 to measure the time interval between the detection of the prompt gamma ray γp and the annihilation gamma ray γa. From this time interval may be derived information regarding the average positron lifetime. This information may be further conditioned and/or processed, if required or desired, by an analyzer 54. Alternatively, other arrangements are possible for determining positron lifetime. For example, in another embodiment, the data processing system 24 could be provided with a high-speed digital oscilloscope with recording capability. One channel of the oscilloscope is connected to the first detector 30, whereas the other channel of the oscilloscope is connected to the second detector 32. Data collected by each channel could then be correlated and analyzed in accordance with the teachings provided herein to determine positron lifetime. However, since systems for detecting positron lifetimes, as well as the algorithms utilized thereby, are well-known in the art and could be easily provided by persons having ordinary skill in the art after having become familiar with the details of the present invention, the positron lifetime algorithm 38, as well as the other systems and detector arrangements that may be required or desired, will not be described in further detail herein. As was briefly mentioned above, the data processing system 24 may also utilize a Doppler-broadening algorithm 40. The Doppler-broadening algorithm 40 assesses the degree of broadening of the 511 keV peak associated with the annihilation gamma rays γa produced by the positron/electron annihilation event. A broadening of the peak is indicative of the presence of one or more lattice defects in the material specimen 12. Such lattice defects may include, without limitation, damage resulting from mechanical and thermal fatigue, embrittlement, annealing, and manufacturing defects. With reference now to FIG. 5, one method for determining the degree of broadening of the 511 keV peak 74 is based on a peak parameter, which may be defined as the number of counts in a central region 76 that contains about half of the total area of the 511 keV peak 74 divided by the total number of counts in the peak. Several different types of Doppler-broadening techniques have been developed and are being used in the positron annihilation art and could be easily implemented in the present invention by persons having ordinary skill in the art after having become familiar with the teachings of the present invention. Therefore, the present invention should not be regarded as limited to any particular Doppler-broadening algorithm. However, by way of example, in one preferred embodiment of the invention, the Doppler-broadening algorithm 40 may comprise the Doppler-broadening algorithm described in U.S. Pat. No. 6,178,218 B1, which is specifically incorporated herein by reference for all that it discloses. With reference now to FIG. 6, the Doppler-broadening algorithm 40 may also involve the use of both of the detectors 30 and 32 comprising the detector assembly 16 in order to determine the degree of broadening of the 511 keV peak 74. For example, in one operational sequence 78, the data processing system 24 monitors one of the detectors (e.g, detector 30) for prompt gamma ray data 20 at step 80. Upon detecting a prompt gamma ray γp, the data processing system 24 then monitors the other of the detectors (e.g., detector 32) and collects positron annihilation data 22 at step 82. After a sufficient amount of positron annihilation data 22 have been collected, the data processing system 24 processes the positron annihilation data 22 in accordance with the Doppler-broadening algorithm 40 at step 84. By way of example, in one preferred embodiment positron annihilation data 22 are collected for a period in the range of about 1-20 nanoseconds (12 nanoseconds preferred) after detecting the prompt gamma ray γp. This method significantly reduces background noise and increases the accuracy of the resulting data. The apparatus 10 of the present invention may be used as follows in order to evaluate a material specimen. A first step in the process involves providing a material specimen 12. The next step of the process involves bombarding the material specimen 12 with neutrons n from the neutron source 14 in order to produce prompt gamma rays γp. This may be accomplished by positioning the material specimen 12 and neutron source 14 adjacent one another so that neutrons n from the neutron source 14 bombard the area or portion of the material specimen 12 that is to be evaluated. In this regard it should be noted that any of a wide range of neutron fluxes and exposure times may be required or desired depending on the particular material specimen 12 to be evaluated. Stated another way, the neutron flux and exposure to the neutron flux should be selected to result in the production of prompt gamma rays γp having energies sufficient to produce a significant number of positrons e+ through the process of pair production. As described herein, prompt gamma rays having energies of at least about 1.1 MeV and preferably about 2.0 MeV, are highly likely to produce positrons by pair production. Accordingly, the present invention should not be regarded as limited to any particular neutron flux or exposure time. However, by way of example, in one embodiment involving a material sample 12 comprising Alcoa 6061/T6 aluminum, a neutron source capable of producing from about 105 to about 106 neutrons per second for ten minutes has been observed to provide sufficient production of prompt gamma rays γp and associated positron annihilation events. Some of the prompt gamma rays γp are emitted from the material specimen 12, whereas others of the prompt gamma rays γp result in the production of positrons e+ in the manner already described. Some of these positrons then annihilate with electrons contained in the material specimen 12, resulting in the production of annihilation gamma rays γa. The emitted prompt gamma rays γp and annihilation gamma rays γa are detected by the detectors 30 and 32. Positron lifetime data are then calculated based on the detected emitted prompt gamma rays γp and the detected emitted annihilation gamma rays γa. The positron lifetime data may then be presented on the display system 26. If the data processing system 24 is provided with a Doppler-broadening algorithm 40, the detected emitted annihilation gamma rays γa are used to produce output data indicative of a lattice characteristic of the material specimen 12. The output data from the Doppler-broadening algorithm 40 may also be presented on the display system 26. A second embodiment 110 of apparatus for evaluating a material specimen 112 is illustrated in FIG. 7. This second embodiment 110 is similar to the first embodiment 10, in that it comprises a neutron source 114 and a detector assembly 116. However, the detector assembly 116 of the second embodiment 110 includes a single detector 130 for detecting both prompt gamma rays γp and annihilation gamma rays γa. As was the case for the first embodiment 10, neutrons n from the neutron source 114 interact with the material specimen 112 to produce prompt gamma rays γp. Some of the prompt gamma rays γp are emitted from the material specimen 112, while others result in the formation of positrons e+ through the process of pair production, illustrated schematically at 118. Many of the positrons e+ produced as a result of the pair production process ultimately annihilate with electrons e− in the material specimen 112, resulting in the formation of annihilation gamma rays γa. The single detector 130 comprising the detector assembly 116 detects both prompt gamma rays γp and annihilation gamma rays γa, and produces prompt gamma ray data 120 and positron annihilation data 122. A data processing system 124 operatively associated with the detector 130 processes the prompt gamma ray data 120 and positron annihilation data 122 in accordance with the methods already described for the first embodiment 10. Thereafter, positron lifetime data and/or information relating to the presence of defects in the material specimen 112 may be presented in human-readable form on a display system 126. Because the data processing system 124 receives both prompt gamma ray data 120 and positron annihilation data 122 from the same detector 130 (as opposed to two different detectors 30 and 32 of the first embodiment 10), the data processing system 124 of the second embodiment 110 is also provided with list mode processing capability in order to process the data based on when it was received, as opposed to integrating the counts at each energy. However, since list mode data processing techniques are well-known in the art and could be easily provided by persons having ordinary skill in the art after having become familiar with the teachings of the present invention, the particular list mode processing technique utilized in the second embodiment 110 will not be described in further detail herein. A third embodiment 210 of apparatus for evaluating a material specimen 212 is illustrated in FIG. 8 and may be used to measure the build-up or accumulation over time of lattice damage that may occur in the specimen 212 during fabrication and/or in-service use. This third embodiment 210 of apparatus for evaluating a material specimen 212 is referred to herein in the alternative as an “on-line sensor” or as apparatus for the “on-line” evaluation of the material specimen 212, in that it is intended to be used to evaluate the material specimen 212 over time. Apparatus for the third embodiment 210 may be similar to either the apparatus for the first or second embodiments 10, 110 described above. That is, the third embodiment 210 may involve the use of two detectors (as is the case for the first embodiment 10) or a single detector (as is the case for the second embodiment 110). Consequently, the third embodiment 210 should not be regarded as being limited to the configuration of either the first embodiment 10 or the second embodiment 110. However, by way of example, as shown and described herein, the third embodiment 210 may comprise a neutron source 214 and a detector assembly 216 having a single detector 230. The neutron source 214 produces neutrons n and directs the neutrons n toward the material specimen 212. Alternatively the neutron source 214 could be replaced with a source of positrons (not shown) having energies of about 3 MeV if the material specimen to be examined is relatively thin. As discussed above, the neutrons n from the neutron source 214 interact with the material specimen 212 and result in the production of prompt gamma rays γp within the material specimen 212. While some of the prompt gamma rays γp are emitted from the material specimen 212, others of the prompt gamma rays γp will result in the formation of positrons e+ within the material specimen 212 through the process of pair production (illustrated schematically at 218). Many of the positrons e+ produced as a result of the pair production process ultimately annihilate with electrons e− within the material specimen 212. The annihilation event results in the formation of annihilation gamma rays γa, most of which are thereafter emitted from the material specimen 212. As was the case for the first and second embodiments 10 and 110, it is generally preferred that the neutrons n from the neutron source 214 have energies in the range of about 0.1 MeV to about 4 MeV. In accordance with this requirement, any of a wide range of neutron sources, such as neutron generators or isotopic neutron sources, may be used in conjunction with the present invention. Examples of neutron generators include, but are not limited to, deuterium-deuterium (D-D) and deuterium-tritium (D-T) generators of the type that are well-known in the art and readily commercially available. An example of an isotopic neutron source includes, but is not limited to, 252Cf. In the embodiment shown in FIG. 8, the neutron source 214 comprises an isotopic neutron source 254, such as, for example 252Cf. In order to accomplish the “on-line” evaluation of the material specimen 212 (e.g., the measurement over time of lattice characteristics of the material specimen 212), it is generally preferred that the neutron source 214 be mounted adjacent to or on the material specimen 212 in a long-term type of arrangement so that neutrons n from the neutron source 214 are directed toward and bombard (i.e., penetrate) the portion of the material specimen 212 that is to be evaluated. For example, if the material specimen 212 to be evaluated comprises a portion of the structure of an aircraft (e.g., a wing spar), the neutron source 214 (e.g., the isotopic neutron source 254) may be mounted or affixed adjacent to the wing spar (i.e., the material specimen 212) to allow the neutron source 214 and detector 216 to be moved or scanned along the wing spar to identify changing conditions in the structure of the wing spar to continually bombard the spar with neutrons. Alternatively, the neutron source 214 may be mounted on the material specimen 212 (e.g., the wing spar) itself. Whether the neutron source 214 is mounted adjacent to or on the material specimen will depend on the configuration (e.g., size) of the material specimen as well as on the environment in which the measurement is to be taken, as would be obvious to persons having ordinary skill in the art after having become familiar with the teachings of the present invention. In addition, the mounting arrangement should be such that the desired portions of the spar (e.g., high stress areas) are exposed to sufficient neutron flux to produce prompt gamma rays γp having sufficient energies to produce a high flux of positrons e+ through the process of pair production in the manner already described herein. Generally speaking, this type of long-term mounting arrangement will be advantageous in a service environment, wherein it is desired to measure lattice damage as it builds-up during the service life, or some portion of the service life, of the material specimen 212. Alternatively, this type of long-term mounting arrangement may be used in other situations, as would be recognized by persons having ordinary skill in the art after having become familiar with the teachings of the present invention. In an alternative arrangement, the neutron source 214 (e.g., the isotopic neutron source 254) could be temporarily placed adjacent to, or on, the material specimen 212 (e.g., a wing spar). Generally speaking, this type of temporary mounting arrangement will be advantageous in a production or fabrication environment, wherein it is desired to measure or monitor lattice damage as it may build-up or be created in the material specimen 212 (e.g., wing spar) during the production process. Alternatively, the temporary mounting arrangement may be used in other situations, as would be recognized by persons having ordinary skill in the art after having become familiar with the teachings of the present invention. As was the case for the other embodiments, any temporary mounting arrangement should be such that the desired portions of the material specimen 212 (e.g., wing spar under manufacture) are exposed to sufficient neutron flux to produce prompt gamma rays γp having sufficient energies to produce a high flux of positrons e+ through the process of pair production in the manner already described herein. Depending on the particular situation in which the third embodiment 210 is utilized, it may be necessary or desirable to provide the neutron source 214 (e.g., the isotopic neutron source 254) with a suitable shield 256 and reflector 258 to reduce stray neutron emission and to help direct additional neutrons n toward the material specimen 212. As was the case for the other embodiments described herein, the shield 256 and reflector 258 may comprise any of a wide range of materials well-known in the art or that may be developed in the future that are or would be suitable for such uses, as would be obvious to persons having ordinary skill in the art after having become familiar with the teachings of the present invention. Consequently, the present invention should not be regarded as limited to a shield 256 and reflector 258 comprising any particular materials. A moderator or thermalizer 260 may also be positioned between the neutron source 214 and the material specimen 212. The thermalizer 260 thermalizes the neutrons n from the neutron source 214, reducing their energies, thereby improving the number of interactions within the material specimen 212. As was described above, the amount of moderation or thermilization to be provided will depend on the energies of the neutrons n from the neutron source 214, as well as on certain characteristics (e.g., thickness, density, etc.) of the material specimen 212 being studied. Generally speaking, it is preferred that the prompt gamma rays γp have energies of at least about 1.1 MeV, and preferably about 2.0 MeV, in order to produce high positron yields through the process of pair production. In one preferred embodiment, the thermalizer 260 comprises a material having a low atomic number, such as polyethylene. The overall length 262 of the polyethylene thermalizer 260 may be changed or varied as necessary to provide the desired degree of moderation or thermalization in accordance with the teachings provided herein. Alternatively, other types of thermalizers comprising other types of materials may be used, as would be obvious to persons having ordinary skill in the art after having become familiar with the teachings of the present invention. It is generally preferred, but not required, to provide additional shielding 264 around the thermalizer 260 in order to further reduce the amount of stray radiation from the neutron source 214 that may reach the detector assembly 216. The presence of such additional shielding 264 will enhance the sensitivity of the detector assembly 216 by reducing the amount of “background” radiation or noise detected by the detector assembly 216. By way of example, in one preferred embodiment, such additional shielding 264 may comprise any of a wide range of bismuth, lead, or borated polymer materials. The detector assembly 216 may be positioned or mounted adjacent the material specimen 212 so that the detector assembly 216 receives both prompt gamma rays γp and annihilation gamma rays γa emitted from the specimen 212. Alternatively, the detector assembly 216 may be mounted to the material specimen 212, again depending on the particular material specimen 212 involved and the environment in which the measurement is to be performed. In the embodiment shown in FIG. 8, the detector assembly 216 comprises a single detector 230 for detecting both prompt gamma rays γp and annihilation gamma rays γa. As was described above for the second embodiment 110, the single detector 230 comprising the detector assembly 216 produces prompt gamma ray data 220 and positron annihilation data 222. A data processing system 224 operatively associated with the detector 230 processes the prompt gamma ray data 220 and the positron annihilation data 222 in accordance with the methods described herein. Thereafter, positron lifetime data and/or information relating to the build-up or accumulation of lattice defects may be captured or “downloaded” from the data processing system, as will be described in greater detail below. As briefly mentioned above, in order to provide for the “on-line” evaluation capability, it is generally preferred that the detector assembly 216 be mounted adjacent to or even on the material specimen 212. Alternatively, the detector assembly 216 could be mounted so that it may be translated along the material specimen 212. The mounting arrangement may be either a long-term mounting system, or a temporary mounting system. In the long-term mounting system, the detector assembly 216 is mounted to (or nearby) the material specimen 212 in order to detect gamma rays over a relatively long term or period. Generally speaking, this type of long-term mounting arrangement will be advantageous in a service environment, wherein it is desired to measure lattice damage as it builds-up or accumulates during the service life, or some portion of the service life of the material specimen 212. Alternatively, this type of long-term mounting arrangement may be used in other situations, as would be recognized by persons having ordinary skill in the art after having become familiar with the teachings of the present invention. In an alternative arrangement, the detector assembly 216 could be temporarily placed on (or nearby) the material specimen 212. Generally speaking, this type of temporary mounting arrangement will be advantageous in a production environment, wherein it is desired to measure or monitor lattice damage as it may build-up or be created in the material specimen 212 during the production or fabrication process. Alternatively, the temporary mounting arrangement may be used in other situations, as would be recognized by persons having ordinary skill in the art after having become familiar with the teachings of the present invention. The data processing system 224 is operatively associated with the detector system 216 and receives the prompt gamma ray data 220 and positron annihilation data 222 produced by the detector system 216. As was briefly described above, the data processing system 224 may process the prompt gamma ray data 220 and positron annihilation data 222 in accordance with a positron lifetime algorithm (e.g., positron lifetime algorithm 38). So processing the prompt gamma ray data 220 and the positron annihilation data 222 results in positron lifetime data. In addition, the data processing system 224 may also process the positron annihilation data 222 in accordance with a Doppler-broadening algorithm (e.g., Doppler-broadening algorithm 40) in the manner already described for the first embodiment 10. Alternatively, if the third embodiment 210 is to be used in a long-term arrangement, such as, for example, to measure the build-up or accumulation of lattice defects during the service life (or some portion of the service life) of the material specimen 212, the data processing system 224 could be simplified considerably to where the data processing system 224 serves merely as a data collection device, collecting the data (e.g., the prompt gamma ray data 220 and/or the positron annihilation data 222) produced by the detector assembly 216, and storing the data in a data storage system 225 for later retrieval and processing in accordance with the teachings provided herein. In such an application, the data storage system 225 could comprise any of a wide range of systems (e.g., magnetic or optical storage media) now known in the art or that may be developed in the future that are or would be suitable for storing such data for such time and under such conditions as may be required by the particular situation. Thereafter, the collected and stored data may be retrieved (i.e., downloaded) and processed in accordance with the teachings herein in order to provide information indicative of the build-up or accumulation of lattice defects during the data collection period. The apparatus 210 of the present invention may be used in accordance with a method 226 illustrated in FIG. 9 to provide for the “on-line” evaluation of a material specimen. A first step 228 in the process involves positioning the neutron source 214 either adjacent to or actually on the material specimen 212. The detector assembly 216 may also be positioned either adjacent to or on the material specimen 212 at step 230. Of course, both the neutron source 214 and detector assembly 216 should be mounted so that neutrons from the neutron source 214 bombard the appropriate or desired portion or portions of the material specimen 212. Similarly, the detector assembly 216 should be mounted so that it will detect gamma rays (e.g., either prompt gamma rays or annihilation gamma rays, or both) emitted by the material specimen 212. The next step 232 of method 226 involves bombarding the material specimen 212 with neutrons n from the neutron source 214 in order to produce prompt gamma rays γp. Again, this may be accomplished by mounting neutron source 214 to the material specimen 212 either in a long-term or a temporary arrangement so that neutrons n from the neutron source 214 bombard the area or portion of the material specimen 212 that is to be evaluated. For example, if the third embodiment 210 is to be used in a long-term situation, such as, for example, to measure the build-up or accumulation of lattice defects during the service life, or some portion of the service life of the material specimen 212, then it will usually be desirable to mount the neutron source 214 on or nearby the specimen 212 so that the neutron source 214 will remain so during that portion of the service life of the specimen 212 that is desired to be monitored. The detector assembly 216 will also be mounted on or nearby the specimen 212 at the appropriate location so that the detector assembly 216 will remain so during the portion of the service life of the specimen 212 that is to be monitored. The next step 234 involves collecting positron annihilation data and/or prompt gamma ray data from the detector assembly 216. The collected positron annihilation data 222 and/or prompt gamma ray data 220 are then processed by the data processing system 224 at step 236 in the manner already described. That is, the positron annihilation data 222 and/or prompt gamma ray data 220 may be processed in accordance with a Doppler-broadening algorithm (e.g., Doppler-broadening algorithm 40) and/or a positron lifetime algorithm (e.g., positron lifetime algorithm 38). The method 226 may then continue to collect and process the positron annihilation data 222 and/or prompt gamma ray data 220 at step 238 over the desired time interval. For example, the desired time interval may be some desired portion of the service life of the material specimen 212 or some desired portion of the production or fabrication sequence involving the material specimen 212. While the data processing system 224 may immediately process the data from the detector assembly 216 in accordance with the description provided herein, the data processing system 224 may also be configured to collect the data storage system 225 the positron annihilation data 222 and/or prompt gamma ray data 220 and store the data on the data storage system 225 for later retrieval and processing. If the third embodiment 210 is to be used in more of a temporary situation, such as, for example, to measure the build-up or accumulation of lattice defects in the material specimen 212 during a more short-term interval, such as during a production or fabrication process that involves the material specimen 212, then the neutron source 214 could be temporarily mounted on or nearby the specimen 212. The neutron source 214 may be left in place during the entire short-term monitoring process, during which time the detector assembly 216 provides prompt gamma ray data 220 and/or annihilation gamma ray data 222 to the data processing system 224. Alternatively, the neutron source 214 may be removed before collecting data from the detector assembly 216. Regardless of whether the neutron source is left in place or removed, the data processing system 224 may process the data 220 and 222 in the manner already described. For example, for monitoring continuous processes, a running average of the response will be processed so that larger pictures of the material specimen 212 can be monitored to assess changes in the material properties that can be fed back to the system controlling manufacture to provide an on-line quality assurance process. Positron lifetime data and/or data indicative of the build-up or accumulation of a lattice defect may then be presented on a suitable display device (e.g., display system 26, FIG. 1). Alternatively, the data may be collected and processed and/or displayed for the user at a later time. It is contemplated that the inventive concepts herein described may be variously otherwise embodied and it is intended that the appended claims be construed to include alternative embodiments of the invention except insofar as limited by the prior art. |
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044366914 | summary | BACKGROUND OF THE INVENTION This invention relates to the confinement of plasmas by magnetic fields and, more particularly, to an improved method and apparatus for the formation of a spheromak plasma (compact torus). Devices employed for the containment of plasmas by magnetic fields may have various configurations. Two well-known types of such devices are the open-ended type, such as the magnetic mirror type, and the toroidal type, such as the tokamak. The underlying principle of all types of such containment devices is the containment of a hot, dense ionized gas away from physical walls for a time sufficient to allow fusion reactions to take place. An advantage of the mirror-type device is that it has a coil-blanket topology which does not link the plasma. However, the mirror-type open ended apparatus has a disadvantage in that since the magnetic field lines do not close upon themselves, the trapped charge particles may escape while travelling along the magnetic field lines which define their spiral orbits. It occurred to many people in the early days of fusion research that mirror end losses could be easily eliminated simply by bringing the two ends of the straight cylinder on themselves, thus forming the well-known torus device. The toroidal-type devices have an advantage in that plasma is well confined in the closed magnetic field lines. Since the ions tend to remain in a spiral orbit about a given set of magnetic field lines the continuity of the magnetic field lines inside the apparatus enhances containment. A tokamak clearly has this above-mentioned advantage but suffers from a difficult topology in which the coil blanket links the toroidal plasma. The spheromak combines the most advantageous aspects of the above-discussed toroidal and mirror schemes. The spheromak is characterized by magnetic field lines which are closed, as in a tokamak, and by a coil blanket topology which does not link the plasma, as in a mirror-type device. Among the advantages of this speromak formation scheme is the ability to keep the physical structure of the apparatus away from the plasma, thus reducing absorbed impurities and keeping the plasma "hot." Also, the spheroidal blanket simplifies the design and construction of the reactor apparatus. The magnetic field configuration of the spheromak includes both toroidal and poloidal components, but the toroidal component is maintained entirely by plasma currents, and, therefore, it vanishes outside the plasma. The outward pressure of the toroidal field and of the plasma is balanced by the inward pressure of a poloidal field. For additional background discussions relating to the spheromak configuration, the reader is referred to S-1 Spheromak, Princeton University, Plasma Physics Laboratory, Aug. 24, 1979, the disclosure of which is hereby incorporated by reference. Three known methods of spheromak plasma formation suitable for spheromak start-up have been experimentally confirmed. The first of these is the so-called "Marshall gun" approach, which is discussed in Alfven, Proceedings of the Second International Conference on Peaceful Uses of Atomic Energy 31 (1958). This approach is characterized by the establishment of an initial poloidal field, followed by the application of toroidal flux through an electrode system. Plasma inertia is relied upon to immobilize the toroidal flux while the poloidal field lines are reconnected within the plasma. This approach has the disadvantage of requiring formation on a dynamic time scale, leading to questions of whether the internal poloidal flux is adequately reconnected. Also, since an electrode system is used, this formation scheme may suffer from problems of erosion and impurity influx, causing plasma cooling problems. Another known scheme suitable for spheromak start-up is the familiar reversed-field theta-pinch approach, as discussed in Centre de Recherches en Physique des Plasmas, Lausanne, Switzerland (1978-79). This scheme is quite similar to the Marshall-gun approach, and thus suffers from the same disabilities. The major difference between the two approaches is that the geometry of the plasma forming structure is rotated by 90.degree. relative to that of the Marshall-gun approach, thus producing radial, rather than axial, plasma acceleration. An improved method and apparatus for inductively forming a detached spheromak plasma configuration wherein the plasma may be contained at a substantial distance away from physical walls is disclosed in a published report entitled S-1 Spheromak, Princeton Plasma Physics Laboratory, Princeton, N.J. (Aug. 24, 1979). The original S-1 Spheromak described in that report is useful for forming a hot plasma, and for generating possibly large quantities of X-rays and neutrons, and can be used in numerous instances where neutrons are needed, as for example in the formation of medical isotopes. The original S-1 Spheromak, which is illustrated in FIG. 1, includes a toroidally-shaped flux core having a radially interior major radius side and a radially exterior major radius side which includes both poloidal and toroidal magnetic field generating coils; a generally spheroidal vacuum vessel for enclosing the flux core; a pair of external equilibrium field coils for supporting the detached plasma; and a pair of pinching coils for pinching off or severing a portion of the plasma and for causing poloidal magnetic field line reconnection, such that the detached plasma may be contained at a distance from physical structure. The original S-1 spheromak has been described in the aforementioned report to operate by energizing the external equilbrium field coils to produce a first poloidal magnetic field; energizing the poloidal coil of the flux core to produce a second poloidal magnetic field, thereby to produce a composite poloidal field which is stronger on the radially exterior major radius side of the flux core than on the radially interior major radius side; energizing the toroidal coil of the flux core to initiate a plasma discharge and to emit toroidal flux which becomes trapped in, and expands, the poloidal flux, such that the plasma expands toward the radially interior major radius side of the flux core, and; pinching off a portion of the distended plasma by energizing the pinching coils so as to produce a detached spheromak plasma. The time variation of the currents applied to each of these coils in accordance this prior method is illustrated in FIG. 2. In U.S. Pat. No. 4,363,776 which is assigned to the same assignee as the present application, the original S-1 Spheromak is described, and its operation to produce a detached spheromak plasma with and without the use of the pinching coil is also described. In accordance with the operation of the original S-1 Spheromak without the use of the pinching coils described in U.S. Pat. No. 4,363,776, the poloidal coil is deenergized at a particular time to produce the spheromak plasma. That is, after the external and poloidal coils have been energized to form the poloidal field and the toroidal coil has been energized to initiate a plasma discharge and to emit toroidal flux which becomes trapped in and expands the poloidal flux, such that the plasma expands toward the major axis of the system, the poloidal coil current is turned off to produce a detached spheromak plasma. Thus, it is known to detach the spheromak plasma by either energizing pinching coils or turning off the poloidal coils at an appropriate time. Experimenters with these two approaches have, however, found that at most only approximately 50 percent of the plasma can be detached from the flux core in this manner. Another limitation of the original S-1 Spheromak is that its arrangement of external equilibrium coils and poloidal field coils resulted in poloidal flux intercepting the flux core and attendant plasma loses to the flux core. SUMMARY OF THE INVENTION The present invention provides a much improved apparatus and method for inductively forming a detached spheromak plasma configuration wherein the detached plasma includes substantially all of the plasma initially surrounding the spheromak flux core. The invention is, therefore, twice as efficient as the prior methods and apparatus and, not using pinching coils, requires only 3 external circuits rather than 4 as in the original S-1 Spheromak described above. The method of the present invention comprises the steps of energizing a set of external equilibrium field coils to produce a first poloidal magnetic field, energizing a poloidal coil formed in a toroidally-shaped ring core with a direct current to produce a second poloidal magnetic field, thereby to produce a composite poloidal field which is stronger on the radially exterior major radius side of the core then on the radially interior major radius side, energizing a toroidal field formed in the ring core to initiate a plasma discharge and to emit toroidal flux which becomes trapped in, and expands the poloidal flux, such that the plasma expands toward the radially interior major radius side of the ring core, and, reversing the direction of the current in the poloidal coil to pinch off most of the distended plasma so as to produce a detached spheromak plasma. The present invention also has an improved flux core. The flux core includes a set of equilibrium coils which serve to reduce the poloidal flux intercepted by the core and thereby reduce plasma loses to the core. The flux core also has a conductive shell which surrounds all of the coils within. The shell serves to stabilize the plasma during the formation phase and causes the surface of the flux core to have a constant poloidal field for time varying fields. |
abstract | A computed tomography scanner is disclosed for performing a spiral scan. In at least one embodiment, the computed tomography scanner includes a rotatable X-ray emitter for generating a beam fan and an X-ray detector, positioned diametrically opposite to the emitter, and an associated evaluation unit. In at least one embodiment, provision is made for an X-ray filter arranged downstream of the X-ray emitter, the position of the X-ray filter being correlated to that of the X-ray detector. Further, the X-ray filter is partly inserted into the beam fan only during operation for generating an unfiltered and, simultaneously, a filtered radiation component of the beam fan, wherein the radiation components have different X-ray spectra and wherein the evaluation unit is designed for separate evaluation of a measurement signal of the unfiltered radiation component and a measurement signal of the filtered radiation component, to obtain dual-energy images. |
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047724463 | summary | CROSS REFERENCE TO RELATED APPLICATIONS Reference is hereby made to the following copending applications dealing with related subject matter and assigned to the assignee of the present invention: 1. "Standardized Reduced Length Burnable Absorber Rods for a Nuclear Reactor" by Barry R. Cooney et al. assigned U.S. Ser. No. 718,902 and filed Apr. 1, 1985. 2. "Burnable Absorber Rod Push Out Attachment Joint" by Joseph B. Mayers et al., assigned U.S. Ser. No. 774,850 and filed Sept. 12, 1985. 3. "Nuclear Reactor Fuel Assembly With a Removable Top Nozzle" by John M. Shallenberger et al., assigned U.S. Ser. No. 644,758 and filed Aug. 27, 1984, now U.S. Pat. No. 4,631,168. 4. "Improved Guide Thimble Captured Locking Tube in a Reconstitutable Fuel Assembly" by Robert K. Gjertsen et al., assigned U.S. Ser. No. 775,208 and filed Sept. 12, 1985, now U.S. Pat. No. 4,684,500. 5. "Burnable Absorber Rod Releasable Latching Structure" by Robert K. Gjertsen, assigned U.S. Ser. No. 807,142 and filed Dec. 10, 1985, now U.S. Pat. No. 4,684,499. BACKGROUND OF THE INVENTION The present invention relates to an apparatus for insertion and removal of releasable burnable absorber rods from the adapter plate of the top nozzle of a nuclear reactor fuel assembly. In a typical nuclear reactor, the reactor core includes a large number of fuel assemblies each of which is composed of top and bottom nozzles with a plurality of elongated transversely spaced guide thimbles extending longitudinally between the nozzles and a plurality of transverse support grids axially spaced along and attached to the guide thimbles. Also, each fuel assembly is composed of a plurality of elongated fuel elements or rods transversely spaced apart from one another and from the guide thimbles and supported by the transverse grids between the top and bottom nozzles. The fuel rods each contain fissile material and are grouped together in an array which is organized so as to provide a neutron flux in the core sufficient to support a high rate of nuclear fission and thus the release of a large amount of energy in the form of heat. A liquid coolant is pumped upwardly through the core in order to extract some of the heat generated in the core for the production of useful work. Since the rate of heat generation in the reactor core is proportional to the nuclear fission rate, and this, in turn, is determined by the neutron flux in the core, control of heat generation at reactor start-up, during its operation and at shutdown is achieved by varying the neutron flux. Generally, this is done by absorbing excess neutrons using control rods which contain neutron absorbing material. The guide thimbles, in addition to being structural elements of fuel assembly, also provide channels for insertion of the neutron absorber control rods within the reactor core. The level of neutron flux and thus the heat output of the core is normally regulated by the movement of the control rods into and from the guide thimbles. Also, it is conventional practice to design an excessive amount of neutron flux into the reactor core at start-up so that as the flux is depleted over the life of the core there will still be sufficient reactivity to sustain core operation over a long period of time. In view of this practice, in some reactor applications burnable absorber or poison rods are inserted within the guide thimbles of some fuel assemblies to assist the control rods in the guide thimbles of other fuel assemblies in maintaining the neutron flux or reactivity of the reactor core relatively constant over its lifetime. The burnable poison rods, like the control rods, contain neutron absorber material. They differ from the control rods mainly in that they are maintained in stationary positions within the guide thimbles during their period of use in the core. The overall advantages to be gained in using burnable poison rods at stationary positions in a nuclear reactor core are described in U.S. Pat. Nos. to Rose (3,361,857) and to Wood (3,510,398). Also, the availability of assemblies of burnable absorber rods on a rapid response basis is required at reactor fuel reload time. The present design of the burnable absorber assemblies, being similar to those illustated and described in the first two patent applications cross-referenced above, includes a plurality of precisely spaced apart absorber rods and thimble plugs fastened at their upper ends to a support plate which also mounts a central hold-down device. In view of the multiplicity of components which make up the absorber assemblies and the precise spacing required between them when they are assembled together, it has been found necessary to assemble the absorber assemblies at a manufacturing facility located remote from the reactor site. The final absorber assemblies are then shipped with the fuel assemblies to the reactor site. This means that the particular absorber assembly design must be specified well in advance of the time of actual reload. A burnable absorber assembly in which the burnable absorber rods have a releasable latching structure is illustrated and described in the fifth patent application cross-referenced above. The advantage of the releasable latching structure is that the configuration of the burnable absorber rods can be specified at the latest possible time because the assembly does not have to include the burnable absorber rods until it is installed. Thus, the nuclear reload design can be fine tuned based on the latest reactor operations input. The ultimate absorber assembly specified may advantageously include, for example, twelve burnable absorber rods and twelve thimble plugs per assembly or other combinations of absorber rods and thimble plugs. Consequently, a need exists for a device that can be used to insert and remove the burnable absorber rods and thimble plugs from the burnable absorber assembly. SUMMARY OF THE INVENTION The present invention provides apparatus for releasably engaging an elongated member, such as a burnable absorber rod or a thimble plug, that is releasably connected to the top nozzle of a nuclear reactor fuel assembly. The top nozzle has an adapter plate disposed at its lower end having at least one passageway therethrough through which the elongated member is disposed. The elongated member has a releasable latching structure at one end that is able to be engaged by the apparatus having at least one latching member movable between a latched position in which the latching member is able to engage said adapter plate and secure the absorber rod in a stationary relationship with respect to the adapter plate and an unlatched position in which the latching member is able to disengage from said adapter plate so that the elongated member be removed from the fuel assembly. The apparatus comprises a hollow releasing member for moving the latching member of the latching structure between its latched position and its unlatched position, an engaging member connected to the releasing member and extending downwardly through the hollow portion of the releasing member, and an actuating member extending downwardly through the hollow portion of the releasing member and coacting with the engaging member to releasably engage the latching structure. More particularly, the present invention enables a plurality of such elongated members to be inserted or removed from a nuclear reactor fuel assembly and comprises a frame and first and second plates disposed within the frame, the second plate disposed below and spaced apart from the first plate. At least one of the first and second plates is capable of vertical movement relative to the other. The apparatus includes means for moving the frame toward and away from the fuel assembly and means for moving the first and second plates within the frame toward and away from the adapter plate of the top nozzle of the fuel assembly. The apparatus thus includes means for varying the vertical distance between the first and second plates between a first distance and a decreased second distance. Means associated with the first and second plates maintains the first and second plates in a position whereby the vertical distance between said plates is the second distance. Further, the apparatus includes a hollow releasing member extending downwardly from the second plate for moving the latching member of the latching structure between its latched position and its unlatched position, an engaging member connected to the releasing member and extending downwardly from the second plate through the hollow portion of the releasing member, and an actuating member extending downwardly from the first plate through the hollow portion of the releasing member and coacting with the engaging member to releasably engage the latching structure when the vertical distance between the first and second plates is the second distance. The present invention coacts with the releasable latching structure that secures the burnable absorber rods and thimble plugs to the top nozzle adapter plate of a nuclear reactor fuel assembly so that the final arrangement of the absorber rods and thimble plugs can be specified at the reactor site. Thus, the latest reactor operating information can be considered when determining the arrangement of the absorber rods and thimble plugs in the design of the fuel assembly. The approach avoids the need for off site manufacturing of the final absorber rod and thimble plug assemblies. Instead, an inventory of individual absorber rods and thimble plugs can be sent to the reactor site prior to fueling the reactor. Once the design of the fuel assembly is specified, the configuration of the absorber rods and thimble plugs can be inserted in the fuel assembly. Later, the spent absorber rods can easily be released from the top nozzle adapter plate and replaced. These and other advantages and attainments of the present invention will become apparent to those skilled in the art upon reading of the following detailed description when taken in conjunction with the drawings wherein there is shown and described an illustrative embodiment of the invention. |
048620076 | abstract | A thermal protection shell for protecting the exterior walls of a radioactive waste container disclosed herein. The shell generally comprises a wall of heat conductive material, such as aluminum or magnesium, which circumscribes and engages the exterior of the waste container walls in intimate, heat-conducting contact under ambient temperature conditions. The thermal coefficient of expansion of the material forming the shell is chosen to be greater than the thermal coefficient of expansion of the material forming the container walls, which are typically steel, so that the heat-conducting contact between the shell and the outer walls is broken when the shell is exposed to a fire. The shell is formed in sections which are rigidly interconnectable by bolt assemblies formed from the same material as the shell itself. The use of such sections allows the shell to be easily mounted over existing radioactive waste containers, and adjusted to fit containers of different diameters. |
041558073 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS As shown in FIG. 1, the core 10 of a nuclear reactor is typically made up of a plurality of elongated fuel assemblies 12. In a typical pressurized water reactor (PWR), for example, the assemblies 12 are supported at the bottom by a lower core plate 14 and are supported at the top by an upper core plate 16. These plates 14, 16 are simple supports for the assemblies 12, generally including a pin-in-hole 17 configuration, which fix axially and radially the location of the upper nozzle 18 and lower nozzle 20 (FIG. 2) of each assembly. The assemblies 12, however, may have radial displacement, during reactor operation, along their length, as a result of flow and temperature induced forces. Prior to reactor operation, the assemblies 12 are loaded into the core 10 closely adjacent one another, and surrounded by a core barrel 22 for former assembly 24. Reactor coolant typically enters the reactor vessel 26 through an inlet nozzle 28, passes down the annular region between the vessel 26 wall and the core barrel 22, is turned 180.degree. in the lower portion of the vessel, passes up through the core 10 at a high velocity, on the order to fifteen feet per second, and exits through the outlet nozzle 30. A typical fuel assembly 12, shown in FIG. 2, comprises a plurality of coextending fuel rods 32 supported at either end by the upper nozzle 18 and lower nozzle 20, and horizontally spaced from one another by spacing means, such as the grids 34 at a plurality of core elevations. The term "core elevation," when used herein and in the appended claims, refers to the axial position within the core that corresponds to the position of a spacing structure or grid, along the assembly length, including the height of the grid outer element, such as a strap 36; for example, if a grid horizontal centerline is positioned fifty inches up from the top of the bottom nozzle, and the height dimension of the outer element of the grid is four inches, then that position between forth-eight and fifty-two inches up from the top of the bottom nozzle is the core elevation. A typical fuel rod 32 includes a plurality of uranium bearing fuel pellets 38 axially stacked within a rod cladding 40 which is hermetically sealed at both extremities by end plugs 42. A typical fuel rod 32 may be as much as sixteen feet in length, or greater, with an outside diameter in the range of one-half inch. The grids 34 function not only to maintain proper lateral alignment and spacing among the rods 32 within a fuel assembly 12, allow for axial growth during reactor heat-up and operation, and control coolant flow, but also to alleviate rod-to-rod contact among fuel rods of the same assembly and also among rods of adjacent fuel assemblies 12. Matched grid core elevations among adjacent assemblies also alleviates concerns regarding coolant cross-flow induced vibration and local flow starvation. It is critical to ensure that any contact among adjacent fuel assemblies 12, which may have a tendency to vibrate and slightly bow, be of a grid-to-grid variety. If such contact is not maintained, there is a likelihood of fretting damage or flow starvation at the point of contact of a rod 32, particularly outermost rods, with another rod 32 or another grid 34. Fretting could possibly lead directly to breach of the cladding 40, and flow starvation could create a local hot spot leading to local cladding 40 melting. Local melting could further lead, in the extreme case, to breach of the cladding 40, thereby allowing reactor coolant to enter the rod, and also allowing the nuclear fuel or fission products to enter the reactor coolant. For these reasons, all the fuel assemblies 12 within a typical core include spacing structures, or grids, at the same core elevations. A typical grid 34 is shown schematically in FIG. 3. As the art of grid and fuel assembly structures is quite crowded, it is to be understood that the instant invention is compatible with most, if not all, fuel assemblies and grid structures in the art, and is not to be limited by the examples discussed herein. A typical grid 34 is an "egg-crate" type structure, manufactured from a plurality of inner straps 44 and outer straps 36 which form a plurality of grid cells 46. Each cell 46 typically includes support means 48 and spring means 50 which support the rod 32 within the cell 46 while allowing some axial and radial expansion. Also typical are components such as mixing vanes 51 and guide tabs 53, among others. During manufacture of one type of fuel assembly 12, a plurality of cylindrical sleeves 52, which typically extend above and below the grid straps 36, 44, are brazed into each grid 34 at given cell locations. The grids are subsequently aligned in a fixture, and thimble tubes 54 inserted through the aligned sleeves 52. A tool is then inserted through each thimble tube 54 which expands the tube 54 at the sleeve 52 locations, thereby fixing the core elevation of each grid 34. Other means of fixing the grid core elevation may be used. The thimble tubes 54 are typically used for insertion of control elements 56, as shown in FIG. 2. As discussed above, all assemblies 12 within a core 10 typically include grids 34 at the same core elevation. However, with increasing technology, there is a likelihood that the number of grids 34, the core elevation of grids 34, or the height of the grids in a given core 10, may desirably be modified. Although the grids 34 are typically composed of a low neutron absorption cross section material, such as alloys of zirconium, they do present a parasitic barrier to nuclear efficiency. It is therefore desirable to minimize the amount of grid material in a core. Also, operating experience may show the desirability of added grids for assembly support and to minimize fuel rod bowing, particularly in the lower portions of the assembly. Similarly, grids may be deleted. It will be well recognized that since a core 10 typically is divided into a plurality of regions into which the assemblies 12 are placed and shuffled during refuelings such that a given assembly may be placed in, for example, three different core locations during its operating lifetime, a significant penalty would result if, because of grid core elevation incompatiblity, assemblies 12 would have to be removed from the core 10 and modified or reprocessed prior to obtaining design burnup. The instant invention provides a core and transition fuel assembly which may be utilized without modifying any other fuel assemblies 12 within the core. The transition assembly provides a spacing structure such as a grid 34, or a grid extension, or a partial grid-like structure, at each core elevation where other fuel assemblies in a given core have a grid 34. The following example is based upon a core 10 which initially includes assemblies 12 with grids 34 at seven core elevations, and is desirably modified to include fuel assemblies 12 with grids 34 at eight core elevations. The basic principle of the instant invention is, however, equally applicable to a decreased number of grids, or merely a change in core elevation of the same number of grids. This example, therefore, is to be viewed as illustrative, and not in a limiting sense. The cored is divided into three core regions, denoted by the letters "A," "B," and "C" on FIG. 4. All three regions initially include seven-grid assemblies when, for example, it becomes desirable to change to assemblies having eight grids. At the next refueling, region "C," which has achieved its desired operating life, is removed for reprocessing. Region "B" fuel assemblies are then shuffled into region "C," and region "A" assemblies are shuffled into region "B." Transition fuel assemblies are then placed in region "A". The transition assemblies are thus placed adjacent region "A" and region "B" assemblies, such that any contact is of a grid-to-grid variety. At the next refueling, region "C" assemblies (seven-grid) are removed, region "B" assemblies (seven-grid) are shuffled to region "C," region "A" transition assemblies are shuffled to region "B," and fresh eight-grid assemblies are placed in region "A." The transition assemblies not maintain contact with both eight-grid and seven-grid assemblies in a grid-to-grid fashion. Dependent upon the core loading pattern, there may be some locations, particularly at the outer portions of the core, where a region "A" assembly (eight-grid) would be adjacent a region "C" assembly (seven-grid), as shown by the dotted line on FIG. 4. At such locations, second transition assemblies would have to be inserted in place of the fresh region "A" (eight-grid) assemblies. The transition assemblies placed in region "B" are subsequently shuffled into region "C," and finally out of the core for reprocessing. The second transition assemblies would subsequently be shuffled into region "B" and then region "A." Upon discharge, the core will the contain only eight-grid assemblies. Because the exact number of assemblies may vary slightly among core regions, there may be instances where one or more assemblies may have to be prematurely discharged, but not premature removal of a whole region or a whole core. One embodiment of a transition assembly for the example seven-grid to eight-grid changeover is schematically shown in FIG. 5. The rectangles represent the core elevations of the grids on the three types of assemblies. In this instance, all of the grid core elevations of the eight-grid assemblies are different than the grid core elevations of the seven-grid assemblies, without overlap. If the grids of the two types are all of the same height, the transition assembly can comprise fifteen grids, also of the same height. If the eight-grid assembly grid height is different from the seven-grid assembly grid height, the transition assembly can comprise seven grids corresponding to the seven-grid assembly, and eight grids corresponding to the eight-grid assembly. It is unlikely, however, that the grid core elevations would not overlap or be within several inches of another at some elevations. FIG. 6 represents these conditions. The eight-grid and seven-grid assembly grids are of the same height, although they may also be different, and the transition assembly comprises nine grids, designed "A" through "I." Grids "A," "D" and "I" may, in this example, be identical to the eight-grid and seven-grid assembly grids in respective core elevation and design. Grids "F," "G," and "H" could also be identical in core elevations with the respective aligned seven-grid assembly grids and eight-grid assembly grids, or as discussed further below, may comprise auxiliary grid structures. Grids "B," "C," and "E" may comprise extended grid structures, also discussed below. FIG. 7 shows a grid structure referred to above as an extended grid structure. As shown, it is generally similar to a typical spacing structure or grid 34, with the addition of vertically extended outer strips 36. A separate piece attached to a standard grid may also be used. The straps are enlarged in the height dimension to maintain grid-to-grid contact among fuel assemblies 12 with grids 34 at relatively close or overlapping core elevations. To minimize the amount of neutron parasitic material in the core, the extended grid may include openings 58. As the extended portion of the outer strap 36, which may be above, below, or above and below the standard portion of the grid, does not necessarily have to provide significant structural ability to the grid, the openings 58 do not detract from necesary structural integrity. It is desirable, however, to orient the openings 58 such that the outermost portions of the outer fuel rods remain enclosed by the extended portion of the outer straps 36. Depending upon the amount of extension necessary, inner butresses 60 may be used to minimize any vibration of the extended portion. The butresses 60 may be separate components, or extensions of the inner straps 44 brazed to the outer straps 36, as shown. FIG. 8 shows a grid structure referred to above as an auxiliary grid structure, as may be used at the locations "F," "G," and "H" of FIG. 6. Here, an auxiliary grid 62 is spaced apart from and affixed to a typical spacing grid 34 by partially extended outer straps 64, which may be integrally formed with the standard grid outer strap 36. They may also be a separate component affixed to the outer strap 36. Partially extended refers to the fact that only a portion of the outer strap 36 is extended upward, or downward, or both, preferably at the strap 36 extremities, thereby minimizing the amount of added parasitic material. Stiffening means 66, such as a punched section or an attached member may be used to increase structural strength. The auxiliary grid 62 further includes at least several auxiliary inner straps 68 to maintain necessary support. Grid-to-grid contact at the auxiliary grid is desirably transmitted to four or more thimble tubes 54 through the partial grid network formed by the straps 68 and through the auxiliary sleeves 70. The sleeves 70 may be brazed to the straps 68. The auxiliary grid may also include partial support straps 71 to provide additional support to the upper horizontal portion 72 of the partially extended outer straps 64. A plurality of inner straps extending across the auxiliary grid 62 would unnecessarily add parasitic material to the core. Other means of affixing the auxiliary grid to the assembly, at a given core elevation, may be utilized. Although the auxiliary grid 62 is shown in FIG. 8 extending above a typical spacing grid, it could alternatively extend above, below, or both above and below the grid. Further, the auxiliary grid 62 may include any of the features typically included in a spacing structure, such as retention means, spring means, mixing vanes, and guide structures, among others. FIG. 9 shows another auxiliary grid structure. In this instance, the auxiliary grid is spaced from and affixed to a typical spacing grid 34, and hence the assembly, by extended sleeves 72. The thimble tubes 54 are expanded at locations inside the extended sleeves 72, which are brazed to the auxiliary inner straps 68. The features of the auxiliary grid 62 may include any of those discussed in reference to FIG. 8, above. An advantage of the structure shown in FIG. 9 is that it eliminates a number of full length inner straps by utilization of the auxiliary straps 60 and partial straps 71. It is therefore seen that the instant invention provides a core and a transition fuel assembly which may be used in a nuclear core with fuel assemblies that include spacing structures or grids at differing, or overlapping, core elevations. Although several example grid structures and transition assemblies have been discussed, it will be apparent that many modifications and variations are possible in view of the above teachings. It is therefore to be understood that within the scope of the appended claims, the invention may be practiced other than as specifically described. |
039881539 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention is directed to a method of manufacturing thin film iris diaphragms for use in corpuscular beam apparatus such as electron microscopes and vidicons which iris diaphragms consist of a metal layer containing at least one opening in order to pass a beam of particles. 2. Prior Art Iris diaphragms have been previously made from two separately manufactured components comprising a metal layer and a substrate or carrier plate. The metal layer is produced by depositing the metal on a profile surface of a template using a vapor metal deposition technique conducted under vacuum conditions. After forming the layer on the template, it is removed and secured to the substrate or carrier plate such as by brazing. Since the rate of fouling of an iris diaphragm increases with the thickness of the metal layer, the metal layer must be as thin as possible. Due to this requirement for a thin layer, the above described process has a high risk of damage to the metal layer as it is being removed from the template and during the step of securing it to the carrier or substrate. In addition thereto, the above described process requires templates which are complex and expensive to manufacture and which can only be used a few times. To overcome the above mentioned problems in the manufacture of iris diaphragms, it has been suggested to produce the diaphragms with the following method. A metal film is deposited directly on a carrier using for example electro-deposition methods with an appropriate mask. The carrier is subsequently etched in the area adjacent to the opening of the thin metal layer by a positive etching technique. Since the etching agent can only attack the unprotected areas of the carrier plate, the metal film or layer and the carrier must be made of different metals. While this process overcomes several difficulties experienced with the previous method, the different in the coefficience of thermal expansion and the difference in the thermal conductivity of the two metals used for the carrier member and the thin film subjects the iris diaphragm to distorsion when used in an apparatus subjected to temperature fluctuations. Thus, the thin film iris diaphragms produced by the proposed method are unsuitable for corpuscular beam apparatuses of the kind in which major temperature fluctuations will occur. SUMMARY OF THE INVENTION The present invention is directed to a method of manufacturing thin film iris diaphragms which have a film as thin as possible and have improved mechanical strength so that they retain their shape even in the presence of substantial temperature fluctuations. In addition, the method of the present invention enables production of iris diaphragms having a complicated structure as well as a plurality of iris diaphragms with minimum manufacturing expenses. To accomplish these tasks, the present invention is directed to a method of manufacturing an iris diaphragm having a thin metal layer with at least one opening therethrough and an integral reinforcing portion of the same metal which is set back from the edge of each of the openings in the thin metal layer with the method comprising the step of providing a surface, forming a first mask on said surface, said mask leaving a portion of the surface exposed with the exposed surface having a configuration of the thin metal layer of the iris diaphragm, applying a thin metal coating to the exposed surface, forming a second mask on the first mask and thin metal layer, said second mask having a configuration of the reinforcing portion and covering a portion of the thin metal layer adjacent the edge of each opening therein, depositing a thick metal layer of the same metal as the thin metal layer to form the reinforcing portion and then removing the iris diaphragm from the surface and the masks. Preferably, the surface on which the first mask is applied is a metal base film applied either directly on a surface of a substrate or onto a bonding layer which has been applied to the substrate. By applying the first mask on the metal base film, the various metal layers may be applied by electrodepositing techniques. Since high resolution can be obtained by a photographic printing operation, each of the masks is preferably formed by utilizing a photo-sensitive material which is exposed and developed to form a mask of the desired configuration. In addition, the use of photographic printing to form the mask enables forming iris diaphragms having complicated shapes. For example ones with extremely small openings or central collectors. The use of photographic techniques also enables forming a plurality of iris diaphragms on the same substrate. The method of the present invention enables the utilization of the same metal for both the thin film and the reinforcing portions so that the temperature fluctuations do not effect the shape of the thin film iris diaphragms. In addition, the process allows producing the thin film or layer to be extremely thin in the neighborhood of the openings so that the thin film iris diaphragm exhibits less tendency to become fouled during operation. Even with this thin metal film or layer of the diaphragm, the mechanical sensitivity is reduced due to the presence of the reinforcing portion so that handling of the thin iris diaphragm no longer presents the problems with regard to damage thereto. |
abstract | Uranium-chelating polypeptides comprising at least one helix-loop-helix calcium-binding (EF-hand) motif which comprises a deletion of at least two amino acid in the 12-amino-acid calcium-binding loop sequence, and their use for uranium biodetection and biodecontamination. |
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claims | 1. A charged-particle multi-beam exposure apparatus (1) for exposure of a target (41) with a plurality of beams of electrically charged particles, said particle beams propagating along parallel beam paths towards the target (41),wherein for each of said particle beams an illumination system (10), a shaping means (20) and a projection optics system (30) are provided, with the illumination system (10) being adapted to produce the beam and form it into a substantially telecentric beam illuminating the shaping means, the shaping means (20) being adapted to form the shape of the illuminating beam into a desired pattern, and the projection optics system (30) being adapted to project an image of the beam shape defined in the shaping means onto the target (40),wherein the shaping means (20) of each particle beam is realized as a pattern definition means for defining a multitude of beamlets in the respective particle beam, said means being adapted to let pass the illuminating beam only through a plurality of apertures defining the shape of beamlets permeating said apertures, said means further comprising a blanking means to switch off the passage of selected beamlets from the respective paths of the beamlets,wherein the illuminating system (10) and/or the projection optics system (30) comprise particle-optical lenses having lens elements (L1, L2, L3, L4, L5) common to more than one particle beam. 2. The apparatus of claim 1, wherein the common lens elements are realized as individual lens elements provided for each of the particle beams and connected to a unique electrical supply. 3. The apparatus of claim 1, wherein the common lens elements are realized by a common structural member surrounding each of the particle beams. 4. The apparatus of claim 1, wherein the particle-optical lenses of the illuminating system (10) and the projection optics system (30) comprise lens elements (L1, L2, L3, L4, L5) common to more than one particle beam as well as lens elements (LB1, LB2, LB3, LB4, LB5) which are individual to one particle beam respectively and connected to individual electrical supplies. 5. The apparatus of claim 1, wherein the particle-optical lenses of the illuminating system (10) and the projection optics system (30) comprise electrostatic lenses (L1, L2, L3, L4, L5) which are common to multiple particle beams as well as electrostatic lens elements (LB1, LB2, LB3, LB4, LB5) which are individual to one particle beam respectively for introducing individual corrections of the effect of the common electrostatic lenses (L1, L2, L3, L4, L5). 6. The apparatus of claim 1, wherein the apertures in the pattern definition means (20) have identical shapes. 7. The apparatus of claim 1, wherein the apertures in the pattern definition means (20) have shapes which produce images of identical shape on the target (41). 8. The apparatus of claim 1, wherein the projection system comprises three or more focusing elements realizing reducing projection optics having two consecutive cross-overs. 9. The apparatus of claim 1, comprising a projection lens system having 4 lenses realizing a 2-stage reduction system, in which parts of the beam are used for beam adjustment and beam analysis at the position of the intermediate image, located between the first and second cross-over. 10. The apparatus of claim 1, comprising a target stage (40) adapted to move the target (41) under the multiple beams according to a predefined scanning motion according to which the beams cover the total area of the target to be exposed in the course of the exposure process. 11. The apparatus of claim 10, wherein the target stage (40) is adapted to perform a scanning motion according to which each beam covers the total area of a sub-field of the target, with the sub-fields altogether totaling to the total area of the target to be exposed. 12. The apparatus of claim 10, wherein the target stage (40) is adapted to perform a scanning motion according to which each beam covers the total area of a sub-field of the target, with the sub-fields of the beams covering separate parts of the total area of the target to be exposed. 13. The apparatus of claim 10, wherein the target stage (40) is adapted to perform a scanning motion according to which each beam covers the total area of a sub-field of the target in a single pass scanning stripe exposure pattern (FIG. 5). 14. The apparatus of claim 1, wherein for each particle beam a pattern definition means is provided, having a pattern field (pf) in which the apertures are located, said pattern field having a length (L) of at least 500 times the size (w) of the apertures. 15. The apparatus of claim 1, wherein for each particle beam a pattern definition means is provided with at least 20000 apertures whose transparency to the particle beam can be electronically controlled between switched on and off states. 16. The apparatus of claim 1, comprising an electrostatic lens (ML) having an electrode column realized as a series of at least 3 electrodes of substantially equal shape of substantially rotational symmetry (EFR, EM, EFN) surrounding the respective beam path, with said electrodes being arranged in consecutive order coaxially along an optical axis representing the center of the beam path and said electrodes being provided with electric supplies for feeding different electrostatic potentials to the respective electrodes. 17. The apparatus of claim 16, wherein the outer radius of all electrodes of the electrostatic lens is not larger than 5 times the largest radius of said particle beam path within the lens. 18. The apparatus of claim 16, wherein the electrodes (EM) of the electrode column are at least partially made from a soft-magnetic material having at environmental conditions a relative permeability greater than 100. 19. The apparatus of claim 18, wherein the relative permeability is greater than 300. 20. The apparatus of claim 16, further comprising a magnetic shielding tube (MS, 3) made from a soft-magnetic material surrounding the electrode column and extending along the direction of the optical axis at least over the length of the electrode column. 21. The apparatus of claim 16, wherein outer portions (OR) of the electrodes (EM) of the electrode column(s) have corresponding opposing surfaces (f1, f2) facing toward the next and previous electrodes, respectively. 22. The apparatus of claim 16, wherein each electrode (EM) of the electrode column comprises an outer member ring (OR) having a cylindrical shape with corresponding opposing surfaces (f1, f2) facing toward the next and previous electrodes, respectively, and further comprises an inner member ring (IR) with a circular edge (cd) directed toward the optical axis. 23. The electrostatic lens of claim 22, wherein the inner member ring (IR) is provided with a concave surface (cv) extending outward from the circular edge (cd) and facing toward the direction from where the charged particles enter the electrode column. |
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description | All nuclear power plants are required to perform visual inspections of specific welds at specified intervals. The conventional method of such a visual inspection attaches a camera on one end of a long pole or rope, and the attached camera is lowered via the pole or rope to an inspection area. Technicians manually deliver these cameras just above the reactor core. This places the technicians in a hazardous environment, as they are exposed to moderate amounts of radiation. Due to the large scope of the required inspections, a high number of technicians are staffed to perform the scope in a minimum time period. This high quantity of personnel is costly for nuclear operators. Related to the different personnel performing these inspections, are the differences in how each technician/inspector performs the specific inspection. These technicians are usually suspended in a large obtrusive platform in or on the water, and this platform interferes with other in-vessel outage activities and can be time consuming and costly to maintain, install and remove. Furthermore, part of the examination process is the requirement to clean the weld before the inspection. The conventional method requires either a brush or hydrolyzing wand to be mounted to an end of a long pole, which is in addition to the pole or rope for the inspection camera. As such, the inspection and cleaning process is a time consuming process, as it requires multiple pole or rope installations and removals to perform a given inspection. Embodiments provide a system and apparatus for visual inspection of a nuclear vessel. The system includes a submersible remotely operated vehicle (SROV) system that is movable to an area within a nuclear vessel. The SROV system includes a maneuverable inspection camera assembly for visual inspection of nuclear vessel components, where the inspection camera assembly is maneuverable in relation to the SROV system. The system also includes a control system located in an area remote from the area within the nuclear vessel. The control system is configured to control the movement of the SROV system and the maneuvering of the inspection camera assembly. In one embodiment, the SROV system includes at least one discharge device attached to the inspection camera assembly, and the at least one discharge device is configured to discharge cleaning fluid to an inspection area. The control system is further configured to control the discharge of the cleaning fluid, and the control system is connected to the SROV system via a control cable. The control system may include a first display unit configured to display the visual inspection performed by the inspection camera assembly, at least one second display unit configured to display a position of the SROV system within the nuclear vessel, a valve system configured to control pressure and flow of the cleaning fluid discharged via the at least one discharge device, and an operational control unit configured to permit a user to control at least one of the movement of the SROV system, the maneuvering of the camera assembly, and the valve system. Further, the operational control unit may b e configured to display tracking information based on the position of the SROV system within the nuclear vessel The SROV system may include a plurality of propulsive devices configured to move the SROV system to the area within the nuclear vessel, and the plurality of propulsive device are remotely controlled by the control system. The plurality of propulsive devices permit the SROV system to move horizontally, vertically, and rotationally about an axis of the SROV system. The SROV system may include a mechanism configured to maneuver the inspection camera assembly. For example, the mechanism is configured to drive a cable that is connected to the inspection camera assembly to move the inspection camera assembly in a vertical direction. The mechanism is also configured to move the inspection camera assembly in a horizontal direction. The SROV system may include a cable retraction mechanism having a pulley to operate the cable and a flotation device connected to the pulley, where the cable retraction mechanism is configured to maintain a tautness of the cable. The cable retraction mechanism may include other components to maintain a tautness of the cable. The SROV system may include a frame assembly having a vertical structure, a n arm assembly connected to the vertical structure driven via a first motor, where the arm assembly is movable about an axis of the first motor, and the first motor is remotely controlled by the control system. In one embodiment, the inspection camera assembly is connected to the arm assembly via a cable and is vertically movable via a second motor that controls movement of the cable, and the second motor is remotely controlled by the control system. The inspection camera assembly may include a camera manipulator having tilt and pan mechanisms, and an inspection camera. The inspection camera is connected to the camera manipulator. The tilt and pan mechanisms permit the inspection camera to be moveable with respect to the camera manipulator, and the camera manipulator is remotely controlled by the control system. The SROV system may include a first positional camera for viewing a first perspective of the SROV system within the nuclear vessel, and a second position camera for viewing a second perspective of the SROV system within the nuclear vessel. The SROV system may include an attachment mechanism configured to attach the SROV system to the area of the nuclear vessel and permit the SROV system to move to different positions on the area. The area of the nuclear vessel may be one of a shroud and Reactor Pressure Vessel (RPV) flange. The system may further include a n observation camera positioned on or near the RPV flange. Also, the SROV system may include an integrated calibration system configured to calibrate an inspection camera of the inspection camera assembly. Embodiments of the present application also provide a submersible remotely operated vehicle (SROV) system for visual inspection of a nuclear reactor. The SROV system includes a device that is movable to an area within a nuclear vessel. The device includes a maneuverable inspection camera assembly for visual inspection of nuclear vessel components. The inspection camera assembly is maneuverable in relation to the device, and the movement of the device and the maneuvering of the inspection camera assembly is remotely controlled. The SROV system may include at least one discharge device attached to the inspection camera assembly, and the at least one discharge device is configured to discharge cleaning fluid to an inspection area. The SROV system may include a plurality of propulsive devices configured to move the device to the area within the nuclear vessel, and the plurality of propulsive devices are remotely controlled. The plurality of propulsive devices permit the device to move horizontally, vertically, and rotationally about an axis of the device. The SROV system may include a mechanism configured to maneuver the inspection camera assembly. The mechanism is configured to drive a cable that is connected to the inspection camera assembly to move the inspection camera assembly in a vertical direction. The mechanism is also configured to move the inspection camera assembly in a horizontal direction. The SROV system may include a cable retraction mechanism having a pulley to operate the cable and a flotation device connected to the pulley, where the cable retraction mechanism is configured to maintain a tautness of the cable. The cable retraction mechanism may include other components to maintain a tautness of the cable. The SROV system may include a frame assembly having a vertical structure, and an arm assembly connected to the vertical structure driven via a first motor, where the arm assembly is movable about an axis of the first motor, and the first motor is remotely controlled. In one embodiment, the inspection camera assembly is connected to the arm assembly via a cable and is vertically movable via a second motor that controls movement of the cable, and the second motor is remotely controlled. The inspection camera assembly may include a camera manipulator having tilt and pan mechanisms, and an inspection camera. The inspection camera is connected to the camera manipulator. The tilt and pan mechanisms permit the inspection camera to be moveable with respect to the camera manipulator, and the camera manipulator is remotely controlled. The SROV system may include a first positional camera for viewing a first perspective of the device within the nuclear vessel and a second position camera for viewing a second perspective of the device within the nuclear vessel. The SROV system may include an attachment mechanism configured to attached the device to the area of the nuclear vessel and permit the device to move to different positions on the area. Also, the SROV system may include an integrated calibration system configured to calibrate an inspection camera of the inspection camera assembly. Various example embodiments will now be described more fully with reference to the accompanying drawings in which some example embodiments are shown. Like numbers refer to like elements throughout the description of the figures. It will be understood that, although the terms first, second, third, etc. may be used herein to describe various elements, these elements should not be limited by these terms. These terms are only used to distinguish one element from another. For example, a first element could be termed a second element, and, similarly, a second element could be termed a first element, without departing from the scope of the embodiments. As used herein, the term “and/or” includes any and all combinations of one or more of the associated listed items. It should be understood that when an element or component is referred to as being “on,” “connected to,” “coupled to,” or “covering” another element or layer, it may be directly on, connected to, coupled to, or covering the other element or layer or intervening elements or components may be present. In contrast, when an element is referred to as being “directly on,” “directly connected to,” or “directly coupled to” another element or layer, there are no intervening elements or layers present. Spatially relative terms (e.g., “beneath,” “below,” “lower,” “above,” “upper,” and the like) may b e used herein for ease of description to describe one element or feature's relationship to another element(s) or feature(s) as illustrated in the figures. It should b e understood that the spatially relative terms are intended to encompass different orientations of the device in use or operation in addition to the orientation depicted in the figures. For example, if the device in the figures is turned over, elements described as “below” or “beneath” other elements or features would then be oriented “above” the other elements or features. Thus, the term “below” may encompass both an orientation of above and below. The device may be otherwise oriented (rotated 90 degrees or at other orientations) and the spatially relative descriptors used herein interpreted accordingly. The singular forms “a,” “an” and “the” are intended to include the plural forms as well, unless the context clearly indicates otherwise. It will be further understood that the terms “comprises,” “comprising,” “includes” and/or “including,” when used herein, specify the presence of stated features, integers, steps, operations, elements and/or components, but do not preclude the presence or addition of one or more other features, integers, steps, operations, elements, components and/or groups thereof. Unless otherwise defined, all terms (including technical and scientific terms) used herein have the same meaning as commonly understood by one of ordinary skill in the art to which the embodiments belong. It will be further understood that terms, e.g., those defined in commonly used dictionaries, should be interpreted as having a meaning that is consistent with their meaning in the context of the relevant art and will not be interpreted in an idealized or overly formal sense unless expressly so defined herein. The visual inspection system and apparatus of the present disclosure provide the capability to remotely manipulate a camera to perform inspections in a nuclear vessel of a reactor. The system and apparatus utilizes a motion control system that allows an operator to be located away from the harsh radioactive environment. For example, the operator uses a control system to maneuver a submersible remotely operated vehicle (SROV) system into a particular area (e.g., shroud or Reactor Pressure Vessel flange) of the nuclear vessel, where a certain component is to be inspected at multiple locations within the reactor vessel. Once the SROV system is in the general location, an inspection camera assembly mounted on the SROV is maneuvered to a more specific location required to perform the inspection. Furthermore, the system of the embodiments provides a cleaning assembly to clean the inspection area. For example, a high-pressure water pump and a series of computer-controlled values are provided at the control system to control the flow and pressure of cleaning fluid to be discharged via a discharge device (e.g., nozzle) mounted on the camera inspection assembly of the SROV system. As such, the SROV system provides a visual overview for the operator while cleaning the inspection area. FIG. 1 illustrates a system 100 for visual inspection of a nuclear reactor according to an embodiment. The system 100 includes a submersible remotely operated vehicle (SROV) system 105 to be operated within a nuclear vessel 101, where the SROV system 105 is connected to a control system 115 via a control cable 109. The control cable 109 includes a plurality of electrical control wires for the transmission of control signals and a minimum of one hose for the transmission of cleaning fluid. The control system 115 is located in an area remote from the area within the nuclear vessel 101. For example, the control system 115 may be located in an area, where the effects of radiation are minimized. The control system 115 is configured to control movement of the SROV system 105, various functions of the SROV system 105 (e.g., a cleaning function), and the maneuvering of specific components within the SROV system 105, by transmitting (and/or receiving) signals to (and/or from) the SROV system 105. These features are further explained below. FIGS. 2A-2D illustrate different views of the SROV system 105 according to an embodiment. In FIG. 2A, the SROV system 105 includes a main device 110 and a cable retraction mechanism 10 6. The cable retraction mechanism 106 is configured to maintain a tautness of a cable 108. For example, the cable retraction mechanism 106, which is submersible in the water of the nuclear vessel 101, maintains the excess cable 108 required for operation from the main device 110. This allows the cable 108 to remain taut. Referring to FIGS. 2B and 2D, one end of the cable 108 is connected to an inspection camera assembly 120 (FIG. 2D) and the other end of the cable 108 is connected to a junction unit 114 via at least one connector 112 (FIG. 2B). When the inspection camera assembly 120 is lowered, the cable retraction mechanism 106 allows the tension of the cable 108 to be maintained. The cable retraction mechanism 106 may include a pulley to operate the cable 108 and a floatation device connected to the pulley. However, the cable retraction mechanism 106 may include any other components, which maintain a tautness of the cable 108. The maneuvering of the inspection camera assembly 120 utilizing the cable 108 is further explained below. Referring to FIGS. 2B-2D, the main device 110 includes a plurality of propulsive devices 113 that are configured to move the SROV system 105 to a certain area within the nuclear vessel 101. The plurality of propulsive devices 113 are remotely controlled by the control system 115. For example, an operator using the control system 115 may direct the SROV system 105 to an area within the nuclear vessel 101 by operating an operational control unit. The details of the control system 115 are further explained with reference to FIG. 6. Based on the operator's instructions, the control system 115 transmits a control signal to the main device 110 of the SROV system 105 via the control cable 109. This control signal controls the amount of thrust each propulsive device 113 generates. Furthermore the control system 115 can be configured to automatically control the amount of thrust each propulsive device 113 generates. This automatic control is initiated when the SROV is commanded to descend to a specified depth or commanded to remain level in the pitch or roll directions. The control cable 109 is connected to the SROV system 105 via at least one connector 112 and a cable support 111. The placement of each propulsive device 113 is further explained below. Referring to FIGS. 2B-D, the main device 110 of the SROV system 105 includes a frame assembly 135, an arm assembly 140 connected to the frame assembly 135, the inspection camera assembly 120 connected to the arm assembly 140 via the cable 108, and an attachment mechanism 130 configured to attached the SROV system 105 to an area of the nuclear vessel 101. The frame assembly 135 includes the junction unit 114, the cable support 111, and the connectors 112. The junction unit 114 provides the distribution of the control signals from the control system 115 via the cable 109. One end of the cable 108 is connected to the junction unit 114 via connectors 112, which are secured by the cable support 111. The frame assembly 135 also includes a first floatation device 116-1 and a second floatation device 116-2, which provide buoyancy to the main device 110 to rise when submerged in the water of the nuclear vessel 101. Also, the frame assembly 135 includes a first propulsive device 113-1, a second propulsive device 113-2, a third propulsive device 113-3, a fourth propulsive device 113-4, a fifth propulsive device 113-5 and a sixth propulsive device 113-6. The first, second, and third propulsive devices 113-1 to 113-3 permit the SROV system 105 to move vertically, and the fourth, fifth and sixth propulsive devices 113-4 to 113-6 permit the SROV system 105 to move horizontally and rotationally. In addition, the frame assembly 135 includes a vertical spine 118, which connects the arm assembly 140 and the attachment mechanism 130 to the frame assembly 135. Further, the sixth propulsive device 113-6 may be connected to the vertical spine 118 to move the SROV system 105 about its axis. For example, the sixth propulsive device 113-6 may be connected on the vertical spine 118 below the arm assembly 140 and above the attachment mechanism 130. As shown in FIGS. 2C and 2D, a first motor 136-1 drives the arm assembly 140 so that the arm assembly 140 is movable about an axis of the first motor 136-1. As a result, the inspection camera assembly 120 can be maneuvered in a more specific location. For example, the arm assembly 140 is connected to the vertical spine 118 via at least one connection point. The first motor 136-1 is provided between one of the connection points and the arm assembly 140 such that the arm assembly 140 pivots about the axis of the first motor 136-1. The first motor 136-1 is controlled by the control system 115. For example, an operator using the control system 115 may direct the arm assembly 140 to a certain position in relation to the axis of the first motor 136-1 by operating an operational control unit. Based on the operator's instructions, the control system 115 transmits a control signal to the main device 110 of the SROV system 105 via the control cable 109. This control signal controls the position of the arm assembly 140. Also, the SROV system 105 includes a mechanism that is configured to drive the cable 108 that is connected to the inspection camera assembly 120 to move the inspection camera assembly 120 in a vertical direction. For example, the inspection camera assembly 120 is connected to the arm assembly 140 via the cable 108. The arm assembly 140 includes a second motor 136-2 that controls movement of the cable 108. For example, based on the control of the second motor 136-2, the cable 108 slides through the upper portion of the arm assembly 140 thereby lowering or raising the inspection camera assembly 120. As a result, the inspection camera assembly 120 is vertically movable. The second motor 136-2 is controlled by the control system 115. For example, an operator using the control system 115 may direct movement of the inspection camera assembly 120 operating an operational control unit. Based on the operator's instructions, the control system 115 transmits a control signal to the main device 110 of the SROV system 105 via the control cable 109. This control signal controls the second motor 136-2 to move the cable 108, thereby moving the camera assembly 120 in a vertical direction. Further, the arm assembly 140 includes a third floatation device 116-3 and a fourth floatation device 116-4, which provide buoyancy to the main device 110 when submerged in the water of the nuclear vessel 101. Also, referring to FIG. 2B, the arm assembly 140 includes a first positional camera 117-1 for viewing a first perspective of the SROV system 105 within the nuclear vessel 101 and a second positional camera 117-2 for viewing a second perspective of the SROV system 105 within the nuclear vessel 101. For example, the first and second positional cameras 117 provide the operator of the SROV system 105 different perspectives on how the SROV system 105 is positioned. The first and second positional cameras 117 are connected to the control cable 109. The main device 110 transmits the positional information from the first and second positional cameras 117 to the control system 115 via the control cable 109. A display device (or multiple display devices) at the control system 115 displays the positional information such that the operator may view the different perspectives of the SROV system 105. This assists the operator to maneuver the SROV system 105 to the correct location. The inspection camera assembly 120 includes a camera manipulator 121 having tilt and pan mechanisms, and an inspection camera 122. The inspection camera 122 acquires visual information regarding the components of the nuclear vessel 101, which are subject to the inspection. The visual information from the inspection camera 122 is transmitted to the control system 115 via a series of internal wires and the control cable 109, and viewed on a display unit of the control system 115. The inspection camera 122 is connected to the camera manipulator 121. The tilt and pan mechanism of the camera manipulator 121 permit the inspection camera 122 to be moveable with respect to the camera manipulator 121. For example, the inspection camera 122 may be tilted upwards and downwards, and rotated around an axis of the camera manipulator 121 via the tilt and pan mechanisms. The camera manipulator 121 is remotely controlled by the control system 115. For example, an operator using the control system 115 may direct movement of the inspection camera 122 in the upward and downward direction and/ or the rotational direction. Based on the operator's instructions, the control system 115 transmits a control signal to the main device 110 of the SROV system 105 via the control cable 109. This control signal controls the camera manipulator 121 to adjust movement of the inspection camera 122. Also, the SROV system 105 includes at least one discharge device 124 (e.g., a nozzle) connected to the inspection camera assembly 120. The at least one discharge device 124 is configured to discharge cleaning fluid to an inspection area. The cleaning fluid is transmitted from a pump and valve system to the at least one discharge device 124 via the cable 109. The control system 115 is configured to control discharge of the cleaning fluid to the inspection area. This feature is further explained with reference to FIG. 6. The attachment mechanism 130 is configured to attach the main device 110 to an area of the nuclear vessel 101 and permit the main device 110 to move to different positions on the area. For example, the attachment mechanism 130 may include a suction device 131 and a minimum of one wheel 132. The suction device 131 permits the SROV system 105 to connect to a surface, and the wheel 132 allows the SROV system 105 to move to different areas of the surface, while the suction device 131 keep the SROV system 105 connected. The operation of the attachment mechanism 130 is controlled by the control system 115. FIGS. 3A and 3B illustrate a portion of the nuclear vessel 101 according to an embodiment. Referring to FIGS. 3A and 3B, the nuclear vessel 101 includes a shroud 190. The nuclear vessel 101 may include other components that are well known to one of ordinary skill in the art. The SROV system 105 may move to a position at the shroud 190 via the plurality of propulsive devices 113 and the positional cameras 117, and attach itself to the shroud 190 via the attachment mechanism 130. As indicated above, the attachment mechanism 130 permits the SROV system 105 to remain attached to the shroud 190, while moving to different positions on the shroud 190, which extends in a radial direction around the nuclear vessel 101. Once the SROV system 105 is attached to the shroud 190, the inspection camera assembly 120 is maneuverable in the manner described above. As such, the operator may view various components surrounding the shroud 190. FIG. 4 illustrates another portion of the nuclear vessel 101 according to an embodiment. Referring to FIG. 4, the nuclear vessel 101 includes a Reactor Pressure Vessel (RPV) flange 191. The SROV system 105 may move to an area at the RPV flange 191 and attach itself to the RPV flange 191, which extends in a radial direction around the nuclear vessel 101. In the same manner, the attachment mechanism 130 permits the SROV system 105 to remain attached to the RPV flange 191, while moving different positions on the RPV flange 191. Further, the embodiments encompass any other area of the nuclear vessel 101, where the SROV system 105 may attach itself for visual inspection of nuclear components. Also, the system 100 may further include an observation camera 175 that is positioned on or near the RPV flange 191. The observation camera 175 provides another perspective o n the positioning of the SROV system 105. For example, the observation camera 175 acquires visual information regarding a n area encompassing the RPV flange 191. This visual information is transmitted to the control system 115 via a camera cable 176. FIG. 5 illustrates an integrated calibration system on the SROV system 105 according to an embodiment. For example, the SROV system 105 may include an integrated calibration system 180 that is configured to calibrate the inspection camera 122. The integrated calibration system 180 includes a plurality of calibration components—first calibration component 180-1, second calibration component 180-2, third calibration component 180-3, and fourth calibration component 180-4. Each of these components provide reference standards visible by the camera 122 to ensure the camera 122 is appropriately calibrated for the examination. FIG. 6 illustrates the control system 115 according to an embodiment. The control system 115 includes a control unit 205, a valve system 210, an operational control unit 215, a plurality of display units 220, and a pump system 225. The control cable 109, in which one end extends to the SROV system 105, is connected to the control unit 205. Further, the control system 115 may include other components that are well known to one of ordinary skill in the art. The operational control unit 215 includes a user interface that allows a n operator to control the discharge of the cleaning fluid, the movement of the SROV system 105 and the maneuvering of the inspection camera assembly 120, as well as any other user controlled function of the system 100. The control unit 205 includes at least one processor and memory storage for controlling operations of the SROV system 105, the valve system 210, and the pump system 225 based on the operator's commands via the operational control unit 215, and for controlling the transmission of visual information from the cameras of the system 100. The valve system 210 includes a plurality of computer controlled operating valves for controlling the flow and pressure of the cleaning fluid, and the pump system 225 includes at least one pump for controlling the amount of cleaning fluid. For example, an operator may use the operational control unit 215 to enter an instruction that controls the flow and pressure of the cleaning fluid. This instruction is received by the control unit 205, and transmitted to the valve system 210 and the pump system 225. The pump system 225 pumps the appropriate amount of cleaning fluid and the valve system 210 controls the flow and pressure of the cleaning fluid, based on the operator's instruction. The cleaning fluid is then provided to the discharge device 124 via the cable 109. Also, the control unit 205 is connected to a plurality of display units 220. A first display unit 220-1 may display visual information from the inspection camera assembly 120, a second display unit 220-2 may display visual information from the first positional camera 117-1, a third display unit 220-3 may display visual information from the second positional camera 117-2, and a fourth display unit may display visual information from the observation camera 175. However, the embodiments of the present application encompass any number of display units 220. For example, the visual information from the cameras may be displayed on one display unit, two display units, or any number of display units. The control unit 205 directs the visual information received via the control cable 109 to the respective display unit 220. Based on the information displayed through the display units 220, the operator may remotely control operation of the SROV system 105. Further, according to one embodiment, the display units 220, the control unit 205, and the operational control unit 215 may be embodied into one or more devices. For example, the control system 115 may include a device such as a touchscreen personal computer (PC). This interface device allows the position of the SROV system 105 to be tracked. For example, the SROV system 105 is positioned in the reactor vessel 101, a reference zero is set, and based on a series of parameters loaded into the control system 115, and the positional feedback information from the SROV system 105 itself, the position of the SROV system 105 is obtained after the SROV system 105 has attached itself to either the RPV flange or the shroud via the attachment mechanism 130. This tracking information may be displayed on the touchscreen PC. As such, the control system 115 allows the ability to track the position of the SROV system 105 once a reference point has been established (e.g., when the main device 110 is attached to a portion of the inspection area via the attachment mechanism 130). In addition, the control system 115 may include a device such as a combo touchscreen PC, in which embodies the display units 220, the control unit 205, and the operational control unit 215 in one unit. However, embodiments of the present application encompass any type of arrangement of the control system 115. Variations of the example embodiments are not to be regarded as a departure from the spirit and scope of the example embodiments, and all such variations as would be apparent to one skilled in the art are intended to be included within the scope of this disclosure. |
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042345553 | claims | 1. The method of decreasing the uranium content of an aqueous HF solution containing uranium to provide a product aqueous HF solution of increased purity, said method comprising: mixing particulate calcium fluoride with said solution to form uranium-bearing particulates, permitting said particulates to sediment from said solution, and separating the resulting aqueous HF solution from the settled particulates. mixing said solution with particulate calcium fluoride to provide a calcium fluoride-to-uranium weight ratio in the range of from about 8 to 75, said calcium fluoride having a nitrogen surface area in the range of from about 1 to 200 m.sup.2 /g, to form uranium-bearing particulates, permitting said particulates to settle out of solution, and selectively recovering the resulting aqueous HF solution. mixing particulate calcium fluoride with said solution to form uranium-bearing particlates, permitting said particulates to sediment from said solution, and separating the resulting aqueous HF solution from the settled particulates. 2. The method of claim 1 wherein said calcium fluoride is mixed with said solution in an amount providing a calcium fluoride-to-uranium weight ratio in the range of from about 8 to 75. 3. The method of claim 1 wherein said calcium fluoride has a nitrogen surface area in the range of from about 1 to 200 m.sup.2 /g. 4. The method of claim 1 wherein said resulting solution is separated by filtration. 5. The method of claim 1 wherein said resulting solution is separated by decanting. 6. The method of decreasing the uranium content of an aqueous HF solution containing a trace amount of uranium, said method comprising: 7. The method of claim 6 wherein said supernate is recovered by decanting. 8. The method of claim 6 wherein said supernate is recovered by filtration. 9. The method of claim 6 wherein the mixing and settling-out operations are conducted at room temperature. 10. The method of decreasing the uranium content of an aqueous HF solution containing uranium without substantially altering the fluoride content thereof, said method comprising: |
claims | 1. A process for the treatment of radioactive graphite, said radioactive graphite including radioactive contaminants, comprising the steps of: reacting radioactive graphite at a temperature above 350xc2x0 C. with superheated steam to form hydrogen, carbon monoxide, and radioactive contaminants; and reacting said hydrogen and said carbon monoxide with oxygen to form water and carbon dioxide in an enclosed vessel under an inert atmosphere; and processing said radioactive contaminants of said radioactive graphite. 2. The process of claim 1 , wherein said process is carried out within the containment of a nuclear reactor. claim 1 3. The process of claim 1 , wherein said second reacting step occurs at a temperature above 250xc2x0 C. claim 1 4. The process of claim 1 , wherein said process further includes the step of converting hydrogen to water using a hydrogen converter. claim 1 5. A process for the treatment of radioactive graphite, said radioactive graphite including radioactive contaminants, comprising the steps of: removing particles of radioactive graphite from a nuclear reactor containment; placing said particles of said radioactive graphite into a fluidized bed reformer; reacting radioactive graphite at a temperature in the range of 250xc2x0 to 900xc2x0 C. with superheated steam to form hydrogen, carbon monoxide, and radioactive contaminants; reacting said hydrogen and said carbon monoxide with oxygen to form water and carbon dioxide in an enclosed vessel under an inert atmosphere; and processing said radioactive contaminants of said radioactive graphite. 6. The process of claim 5 , wherein said process further includes the step of reducing the size of said particles of said radioactive graphite prior to placing said particles into said fluidized bed reformer. claim 5 7. The process of claim 6 , wherein said reducing step includes reducing said particles of said graphite to a size less than 12.0 cm. claim 6 8. The process of claim 5 , wherein said placing step includes placing said particles of said radioactive graphite into said fluidized bed reformer with a screw conveyor. claim 5 9. The process of claim 5 , wherein said process further includes the step of introducing said particles of said radioactive graphite and water into a size-reduction grinder to form a slurry that is introduced into said fluidized bed reformer with an injector pump. claim 5 10. The process of claim 5 , wherein said process includes reacting said radioactive graphite at a temperature in the range of 600xc2x0 C. to 700xc2x0 C. claim 5 11. The process of claim 5 , wherein said process includes removing said radioactive contaminants with a high temperature filter. claim 5 12. The process of claim 5 , wherein said process includes removing said radioactive contaminants with a wet scrubber. claim 5 13. The process of claim 5 , wherein said first reacting step includes adding said oxygen to said superheated steam to provide energy to form said carbon monoxide, said hydrogen, and said radioactive contaminants from said radioactive graphite. claim 5 |
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abstract | A method of operating a nuclear reactor is disclosed. The reactor (1) encloses a core having a plurality of fuel rods (9). Each fuel rod (9) includes a cladding and fuel pellets of a nuclear fuel. The fuel pellets are arranged in an inner space of the cladding leaving a free volume comprising an upper plenum, a lower plenum and a pellet-cladding gap. The reactor is operated at a normal power and a normal inlet sub-cooling during a normal state. The reactor is monitored for detecting a defect on the cladding of any of the fuel rods. The operation of the reactor is changed to a particular state after detecting such a defect. The particular state permits an increase of the free volume in the defect fuel rod. The reactor is operated at the particular state during a limited time period, after which the reactor is operated at the normal state. |
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047598972 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT Referring to FIG. 1, there is illustrated a nuclear reactor installation 20, including a reactor core 21, interconnected by a transfer canal 22 with a spent fuel pool 23, which is sunk beneath the deck 24 of a spent fuel building (not shown). The spent fuel pool 23 is generally rectangular in shape, comprising four vertical walls 25 and filled with water to a level 28 (FIG. 2). Mounted in the spent fuel pool 23 are a plurality of storage racks 26 (only several illustrated). Typically, there are disposed in the reactor core 21 between 160 and 190 nuclear fuel assemblies 30 (see FIG. 3). Periodically, during refueling operations, spent fuel assemblies 30 are removed from the reactor core 21 and the remaining fuel assemblies 30 are rearranged within the reactor core 21, fresh, unused fuel assemblies 30 being added around the outer perimeter of the reactor core 21 and partially spent fuel assemblies 30 being moved radially inwardly toward the center. Typically, after about three such refueling cycles, a nuclear fuel assembly 30 will reach the center of the reactor core 21 and be completely spent. Spent fuel assemblies 30 are removed through the transfer canal 22 to the spent fuel pool 23 by transport means (not shown). Also, partially spent fuel assemblies 30 may be moved to the spent fuel pool 23 for service procedures, including measurements of the kind for which the present invention is used. An upender 27 shifts the fuel assembly 30 from its vertical orientation in the reactor core 21 to a horizontal orientation for transport along the transfer canal 22. The fuel assembly 30 is then again shifted to a vertical position and transferred by an overhead handling crane (not shown) to the spent fuel pool 23, the complete fuel assembly 30 being deposited in one of the storage racks 26, all in a known manner. Referring to FIG. 3, a typical fuel assembly 30 includes a cluster of approximately 200 elongated, parallel fuel rods 31, arranged in a substantially square configuration and extending between a top nozzle 33 and a bottom nozzle 34, each generally in the form of a rectangular block. The entire fuel assembly 30 is approximately 160 inches in overall length. The proper spacing of the fuel rods 31 in the rectangular configuration, is maintained by a plurality of spaced-apart grids 35, which may typically be spaced about two feet apart. Each of the nozzles 33 and 34 and the grids 35 is substantially square in transverse cross section, the nozzles 33 and 34 each being provided with four substantially planar external faces 36 arranged to be substantially parallel, respectively, with the four sides of each of the grids 35. The bottom nozzle 34 has four depending support pedestals 37, two of which contain locating receptacles (not shown) for respectively receiving locating pins in the reactor core 21, accurately to position the fuel assembly 30 therein. Referring to FIGS. 16 and 17, the fuel assembly 30 may, in use, undergo dimensional distortion which may be of several types, including "bow", "twist" and "tilt". More particularly, if the top nozzle 33 is arranged with its faces 36 vertical, the tilt of the fuel assembly 30 is defined as the distance from the centroid 142 of the bottom nozzle 34 to a vertical line 140 passing through the centroid 141 of the top nozzle 33. The centroids 141 and 142 are approximated by the centers of squares defined by the intersection of the faces 36 of the top and bottom nozzles 33 and 34 with horizontal planes passing therethrough. The bow of the fuel assembly 30 at a grid 35 is defined as a deviation of the centroid 144 of the grid 35 from a straight line 143 connecting the centroids 141 and 142 of the top and bottom nozzles 33 and 34. The centroid 144 of a grid 35 is approximated by the center of a square defined by the intersection of the side faces of the grid 34, when oriented vertically with a horizontal plane passing therethrough. The twists of the fuel assembly 30 at a grid 35 is defined as the angular rotation of the grid 35 from a reference plane defined by a face 36 of the top nozzle 33. Bow and tilt are measured in inches and the twist angle is measured in degrees. Referring to FIG. 2 there is illustrated a measurement system, generally designated by the numeral 40, constructed in accordance with and embodying the features of the present invention, for determining the bow, twist and tilt of a nuclear fuel assembly. The measurement system 40 includes a control and processing unit 41 which is disposed outside the spent fuel pool 23, preferably remote therefrom, and an underwater measurement assembly 50 which is mounted on the deck 24 of the spent fuel pool 23, and which extends in use below the water level 28 in the spent fuel pool 23. The control and processing unit 41 includes an isolation transformer 42 coupled to an associated source of AC power and to a computer 43, a printer 44 and a control console 45, the computer 43 also being connected to the printer 44 and the control console 45, the latter being in turn connected to a hydraulics panel 46 for controlling the hydraulic operation of the measurement system 40. The underwater measurement assembly 50 includes an upper junction box 118 which is connected by cabling to the control console 45 and by tubing to the hydraulics panel 46, the upper junction box 118 being mounted in a fixed position. The upper junction box 118 is connected to a lower junction box 124 mounted on movable parts of the underwater measurement assembly 50. The upper junction box 118 is also electrically connected to a motor 112, a potentiometer 117 and two limit switches 119, these parts all being mounted in fixed positions. The lower junction box 124 is connected to four movable measuring gauges 135. The underwater measurement assembly 50 is the portion of the measurement system 40 which is submerged in the spent fuel pool 23 or the transfer canal 22 and in which the spent fuel assembly 30 is positioned for measurement. The control and processing unit 41 is a computer-based unit, the computer being used to control the sequence of operations and to perform the calculations required to convert the raw data from the measurement gauges 135 into bow, twist and tilt form. The use of the computer 43 allows for a fully automatic mode of operation which can significantly reduce required measurement times and operator fatigue. In addition, the computer 43 can be bypassed to allow manual control of system functions. The control and processing unit 41 is programmable to accommodate differences among types of fuel assemblies. Referring now to FIGS. 4-15, the underwater measurement assembly 50 includes a support assembly 51 which is coupled by a mast assembly 70 to a frame 80, on which are mounted a lower clamp assembly 95 and an upper clamp assembly 100 (FIG. 10) for positioning a fuel assembly 30 with respect to the frame 80. A drive assembly 110 is coupled to the frame 80 and drives a carriage assembly 120 along the frame 80, longitudinally of the associated fuel assembly 30, the carriage assembly 120 carrying a pair of measuring units 130 for effecting the required measurements. Referring in particular to FIGS. 4-7, the support assembly 51 includes a flat, rectangular base plate 52 which is adapted to be disposed on the deck 24 of the spent fuel pool 23, the forward edge of the base plate 52 preferably being disposed substantially flush with the upper edge of one of the walls 25 of the spent fuel pool 23. Integral with the rear end of the base plate 52 and projecting upwardly therefrom are one or more threaded studs 53, adapted to be received through the complementary openings in a plurality of counterweights 54, which may be stacked on the base plate 52 and secured in place by suitable nuts engaged with the studs 53 to hold the support assembly 51 in place. Overlying the base plate 52, substantially parallel thereto, is a generally triangular support plate 55 provided with a plurality of spaced-apart support pads 56 depending therefrom for engagement with the base plate 52. Formed adjacent to the apex of the support plate 55 is an elongated slot 57, for a purpose to be explained more fully below. Integral with the forward end of the support plate 55 are two laterally spaced-apart, upstanding sockets 58 in which are respectively threadedly received two studs 59. A mounting bar 60 spans the sockets 58, the bar 60 being provided at its opposite ends with end blocks 61 having bores therethrough for respectively receiving the upper ends of the studs 59. Adjusting nuts 62 being provided on the studs 59 to move the studs 59 up or down for thereby adjusting the inclination of the bar 60. The support assembly 51 also includes a fore and aft adjusting screw 63, threadedly engaged in a lug 64 and bearing against the rear end of the support plate 55. A screw 65 extends through the slot 57 and is threaded into the base plate 52. The lug 64 is fixed to the base plate 52. Thus, by rotation of the screw 63, the fore and aft position of the support plate 55 may be set, after which the screw 65 is tightened to lock the support plate 55 against forward movement. There are also provided a pair of side adjusting screws 66, respectively coupled at their forward ends by clevis links 67 to side edges of the support plate 55 by pivot pins 68 (FIG. 7). The screws 66 are respectively threaded in lugs 69 which are fixed to the base plate 52. Thus, it will be appreciated that the screws 66 may be utilized to effect lateral or side-to-side adjustment of the position of the support plate 55. In use, the support assembly 51 is preferably arranged so that the forward end of the support plate 55 projects beyond the forward end of the base plate 52 over the spent fuel pool 23, the orientation of the support plate 55 and the mounting bar 60 being adjusted so that the mounting bar 60 is disposed substantially horizontally and substantially parallel to the adjacent wall 25 of the spent fuel pool 23. The mast assembly 70 includes a pair of laterally spaced-apart, vertical plates 71 interconnected at the forward and rearward ends thereof by braces 72. Formed in the underside of the plates 71 adjacent to the rearward ends thereof are hooks (not shown) adapted to fit over the mounting bar 60 for supporting the mast assembly 70 in cantilever fashion. Integral with the plates 71 at their forward ends is a cylindrical upper mast section 74, which depends from the plates 71 and may be coupled to a mast extension 75, which may be used to provide additional length to the mast assembly 70, if desired for particular applications. The upper mast section 74 and the mast extension 75 are both provided at their lower ends with a mounting flange 76, adapted to be fixedly secured to the frame 80 for supporting same beneath the water level 28 of the spent fuel pool 23. Referring now to FIGS. 4, 5 and 10, the frame 80 includes a top plate 81, to the rear end of which the bottom of the mast assembly 70 is fixedly secured. The forward end of the top plate 81 is provided with a cutout 82 which defines a pair of arms 83. Fixedly secured to the underside of the top plate 81 parallel thereto is a flat mounting plate 84 (FIG. 4), which is in turn fixedly secured to the upper end of an elongated strongback 85. The strongback 85 is channel-shaped in transverse cross section, opening rearwardly toward the adjacent wall 25 of the spent fuel pool 23, extends substantially perpendicular to the top plate 81, and has a length substantially greater than the overall length of an associated fuel assembly 30. Spanning the legs of the strongback 85 and fixedly secured thereto at the upper end thereof is a rectangular plate 86 to which are fixedly secured a pair of laterally spaced apart, and rearwardly extending gussets 87. Each of the gussets 87 has formed in its bottom edge a hook 88. In use, the hooks 88 may be mounted directly on the mounting bar 60, in certain applications to aid in installation of the frame 80. Preferably, a stiffener bar 89, generally T-shaped in transverse cross section, is secured to the rear wall of the strongback 85 between the legs thereof for reenforcement purposes (FIG. 10). Integral with the strongback 85 and projecting forwardly therefrom are a pair of laterally spaced-apart webs 90 each integral at its distal edge with an elongated rail 91 (FIGS. 10 and 13). The rails 91 are parallel and extend longitudinally of the strongback 85 a substantial portion of the length thereof. Fixedly secured to the strongback 85 adjacent to the lower end thereof and projecting rearwardly therefrom is a standoff assembly 92, adapted to engage the adjacent wall 25 of the spent fuel pool 23 and being adjustable to cooperate with the support assembly 51 to support the frame 80 in a use orientation with the strongback 85 disposed substantially vertically. Fixedly secured to the strongback 85 adjacent to the lower end thereof and projecting forwardly therefrom is a support bracket 93 on which is mounted the lower clamp assembly 95. More particularly, the lower clamp assembly 95 may include an upstanding pedestal 96 which carries at its upper end a hydraulic cylinder 97, the piston rod 98 of which projects vertically upwardly and carries at its upper end a support pad 99 adapted to engage the bottom nozzle 34 of an associated fuel assembly 30. Respectively mounted on the arms 83 of the top plate 81 are the two upper clamp assemblies 100, which are substantially identical in construction, wherefore only one will be described in detail. Each of the upper clamp assemblies 100 includes a mounting bracket 101 which is fixedly secured to the arm 83 and depends therefrom. Mounted on the mounting bracket 101 is a hydraulic cylinder 102 which is disposed with the piston rod 103 thereof projecting horizontally forwardly into the cutout 82 between the arms 83 of the top plate 81. Fixedly secured to the piston rod 103 at its distal end is a clamping jaw 105 provided with a V-shaped notch 106 therein. Also fixedly secured to the clamping jaw 105 are the forward ends of a pair of guide rods 107 which extend horizontally through complementary openings in the mounting bracket 101. The hydraulic cylinders 102 and 97 are coupled by suitable conduits (not shown) to a hydraulic manifold 108 (FIG. 5) mounted on the top plate 81, and in turn coupled through the remote hydraulic panel 46 to an associated source of pressurized hydraulic fluid. In operation, the associated fuel assembly 30 to be measured is moved by means of its overhead handling crane into position with the top nozzle 33 in the cutout 82 of the top plate 81, and the cylinders 102 are operated to move the clamping jaws 105 respectively into clamping engagement with opposite corners of the top nozzle 33. Thus, it will be appreciated, as can be seen in FIG. 5, that the V-shaped notch 106 of each clamping jaw 105 engages adjacent faces 36 of the top nozzle 33, the support assembly 51 being arranged so that when the fuel assembly 30 is thus clamped in position, the faces 36 of the top nozzle 33 are arranged substantially vertically. When the top nozzle 33 has been thus clamped in position, the lower clamp assembly 95 is elevated by operation of the cylinder 97 to bring the support pad 99 into engagement with the bottom nozzle 34. Thus, the upper and lower clamp assemblies 100 and 95 cooperate fixedly to position the fuel assembly 30 in a substantially vertical orientation for measurement. It will be appreciated that, at all times, the fuel assembly 30 is supported by its associated overhead handling crane, the clamp assemblies 95 and 100 serving only to laterally position the fuel assembly 30 and hold it in its predetermined measurement orientation against any lateral forces which might be applied by the measuring apparatus, as described below. The drive assembly 110 is carried by the top plate 81, and includes an upstanding bracket 111 to which is fixedly secured the electric drive motor 112. Secured to the output shaft of the motor 112 is a sprocket 113 which engages a drive chain 114 which extends downwardly through a complementary opening in the top plate 81 in front of the strongback 85, the chain 114 being engaged with a bottom sprocket 115, carried by the strongback 85 immediately above the support bracket 93 (FIGS. 14 and 15). The output shaft of the motor 112 is also coupled by a sprocket and chain coupling 116 to a rotary potentiometer 117 which is mounted on the top plate 81. The electrical conductors (not shown) for the motor 112 and the potentiometer 117 are coupled through the junction box 118 mounted on the top plate 81, and thence to the remote control console 45. Also carried by the top plate 81 and depending therefrom along one side edge thereof is a limit switch 119 (FIGS. 5 and 8), a similar switch (not shown) being provided near the lower end of the frame 80 for a purpose to be explained below. Referring in particular to FIGS. 4 and 10-13, the carriage assembly 120 is fixedly secured to the drive chain 114 by a suitable coupling means (not shown). The carriage assembly 120 includes a mounting bracket 121 provided with two rearwardly projecting bearings 122, which are respectively disposed in sliding engagement with the rails 91 for guiding vertical movement of the carriage assembly 120 by means of the drive chain 114. Fixedly secured to the mounting bracket 121 are two mutually perpendicular rectangular support plates 123, disposed substantially perpendicular to the top plate 81 so as to be respectively parallel to the inner two faces 36 of the top nozzle 33 of the fuel assembly 30 when it is clamped in its measurement position in the frame 80, as illustrated in FIGS. 4, 5 and 10. Carried by the mounting bracket 121 is the junction box 124 which is electrically connected to the junction box 118. Mounted on one of the support plates 123 and depending therefrom is a light 125 for illuminating the region of the fuel assembly 30 being measured, so that it can be viewed on an underwater closed circuit TV camera 126, which is fixedly secured on the same support plate 123 (see FIGS. 10-12). Fixedly secured to the lower clamp assembly 95 and projecting rearwardly therefrom at substantially the level of the lower end of the fuel assembly 30 when it is mounted in its measurement position, is a check plate 127 which is used to verify the repeatability of the measurement tests conducted with the measurement system 40. Respectively carried by the suport plates 123 are two measuring units 130 (FIG. 10) which are substantially identical in construction, wherefore only one will be described in detail. Each of the measuring units 130 includes a support bracket 131 which is coupled to the support plate 123 and projects rearwardly therefrom. Also fixedly secured to the support plate 123 and projecting rearwardly therefrom are a pair of parallel guide rods 132 which extend through the complementary openings in a slide plate 133 for guiding the sliding movement thereof. A hydraulic cylinder 134 is fixedly secured to the support bracket 131, and has a piston rod 134a (FIG. 11) which is fixedly secured to the slide plate 133 for effecting fore and aft reciprocating movement thereof. Carried by the slide plate 133 and projecting forwardly therefrom are a pair of laterally spaced-apart, elongated measurement gauges 135, respectively extending forwardly through complementary openings in the support plate 123. Carried by the support plate 123 and projecting rearwardly therefrom is a stop member 136 for limiting the forward movement of the slide plate 133. Preferably, each of the measurement gauges 135 is of the linear variable differential transformer (LVDT) type. A LVDT is an electromechanical transducer that produces an electrical output proportional to the displacement of a separate movable core. As the slide plate 133 is moved forwardly by the cylinder 134, the movable cores of the LVDT gauges, which project forwardly therefrom, engage the adjacent face of a grid 35 or nozzle 33 or 34 of the fuel assembly 30, and will be displaced or retracted relative to the supporting sleeve of the gauge 135 unil the slide plate 133 stops against the stop member 136. The gauge 135 produces an output signal which is proportional to the amount of retraction of the movable core thereof, which is in turn proportional to the distance of the face being measured from an associated reference plane parallel thereto. The carriage assembly 120 also includes a cable weight roller 137 (FIG. 13) which is carried by a bracket 138 provided with bearings 139 disposed in sliding engagement with the rails 91 of the frame 80. In use, the hydraulic and electrical cables from the upper junction box 118 and associated hydraulic manifold 108 extend downwardly around the roller 137 and up to the junction box 124 and the hydraulic cylinders 134 on the carriage assembly 120. The cable weight roller 137 hangs freely and keeps the cables taut so as to prevent any tangling or interference with the vertical movement of the carriage assembly 120. Referring now to FIGS. 4, 10 and 18, the operation of the measurement system 40 will be explained in detail. First of all, the underwater measurement assembly 50 is mounted in place and adjusted so that the strongback 85 is diposed substantially vertically and the top plate 81 is disposed substantially horizontally. Once the underwater measurement assembly 50 has thus been accurately positioned in the spent fuel pool 23, all further operations can be effected under remote control. First of all, the fuel assembly 30 is positioned in the underwater measurement assembly 50 in the measurement orientation, illustrated in FIGS. 4, 5 and 10, with the top nozzle 33 disposed in the cutout 82 of the top plate 81, and with the fuel assembly 30 hanging freely from its overhead handling crane. The fuel assembly 30 is rotated until opposite corners thereof face the arms 83 of the top plate 81, whereupon the upper clamp assemblies 100 are actuated to bring the clamping jaws 105 into engagement with the adjacent corners of the top nozzle 33 for preventing lateral movement thereof. Then the lower clamp assembly 95 is actuated to bring the support pad 99 up into engagement with the bottom nozzle 34 with a force sufficient to prevent sway of the fuel assembly 30 as a result of any lateral loads which might be applied by the measurement gauges 135. Then, the carriage assembly 120 is moved to its uppermost or zero position illustrated in FIG. 4, that position being set to correspond to a predetermined output reading from the potentiometer 117. The upper limit switch 119 stops the carriage assembly 120 in the event it overruns the zero position. In this uppermost position, the carriage assembly 120 is disposed opposite the top nozzle 33. It is a significant aspect of the invention that when the underwater measurement assembly 50 is disposed in its measurement position illustrated in the drawings, the measurement gauges 135 are all disposed in a common horizontal plane which, in the uppermost position of the carriage assembly 120 passes through the top nozzle 33 intermediate the upper and lower ends thereof. Next, the measurement gauges 135 are advanced against adjacent faces 36 of the top nozzle 33, moving into engagement therewith at four points which define two lines intersecting in a horizontal plane at a corner of the top nozzle 33. The measurement data from the gauges 135 are transmitted through the control console 45 to the computer 43 and are printed out by the printer 44. The carriage assembly 120 is then moved downwardly along the fuel assembly 30, stopping at each grid 35 and at the bottom nozzle 34 at positions determined from the potentiometer 117 readings, to repeat the measurement process. At each of these measurement locations, the carriage assembly 120 will be so positioned that the plane of the gauges 135 passes through the associated grid 35 or bottom nozzle 34 intermediate the upper and lower ends thereof, the measurement data taken being printed out at each stage. At the bottom of the travel of the carriage assembly 120, the gauges 135 are advanced against the check plate 127 to verify the performance and repeatability of the measurements of the gauges 135. The bottom limit switch is actuated in the event of overrun of this check position. The computer 43 then processes the raw data and calculates the bow, twist and tilt of the fuel assembly 30, and this information is then printed out by the printer 44. More specificaly, the computer 43 utilizes the measurement points to calculate two line equations which define lines lying along the adjacent faces of the nozzle or grid, and also to calculate the intersections of these lines which corresponds to a corner of the grid or nozzle. From this information the square defined by the intersection of the grid or nozzle with the measurement plane of the gauges 135 is determined, and vector techniques are used to compute the center of that square and its angular orientation with respect to reference coordinates. The details of this process may better be understood by reference to FIG. 18. Programmed into the computer are orthogonal X, Y and Z coordinates, wherein the Z axis is parallel to the strongback 85, i.e., vertical in this case or perpendicular to the plane of the paper in FIG. 18. Thus, it will be appreciated that the XY plane is horizontal, being defined by the gauges 135, while the YZ and XZ planes are vertical, XZ plane perferably being defined by the strongback 85. The points A, B, C and D represent the points of engagement of the measurement gauges 135 with the faces of the grid or nozzle being measured. The output signals from the gauges 135 are proportional to the distances of the points A and B from the Y axis and the distances of the points C and D from the X axis. The gauges 135 corresponding to the points A and B are parallel to the X axis and spaced therefrom by distances b and a, respectively. Similarly, the gauges 135 corresonding to the points C and D are parallel to the Y axis and spaced therefrom by distances a and b, respectively. Since the distances a and b are known and programmed into the computer 43, this raw data provides the three-dimensional coordinates of the points A, B, C and D and permits calculation of the equations of the lines AB and CD and the extension thereof to their intersection at the point I, at a corner of the grid or nozzle being measured. Since the lengths of the sides of the nozzle or grid being measured are known, the coordinates of the other corners G, H and J thereof can be calculated. Next, the angle W that the line IG makes with the X axis is calculated, which will be equal to the angle V that the line IH makes with the Y axis. Then, the magnitude and direction of the vector OE is computed, and the vector OE is resolved into its X and Y components to identify the center point E of the square IHJG. This center E approximates the centroid of the grid or nozzle being measured. Once the coordinates of the center E has been determined for each of the grids and nozzles 33-35, it is a simple matter to calculate the lines 140 and 143 in FIG. 16 and to determine the tilt of the fuel assembly 30 and the bow and twist thereof at each grid 35. In this regard, the angle W of inclination of the top nozzle 33 constitutes a zero reference and the corresponding angles for each of the other grids 35 and the bottom nozzles 34 are compared to this reference to give a measurement of the twist of the fuel assembly 30 at each such grid or nozzle. After the measurement calculation operation is complete, the carriage assembly 120 is returned to its upper position in preparation for reception of the next fuel assembly 30 to be measured. The foregoing operations and calculations are performed under the control of a computer program, the flow chart for which is set forth in FIGS. 19-27. Referring now to FIGS. 19-27, the operation of the system 40 will be explained in connection with the program for the computer 43. The program, which is generally designated by the numeral 150, is a menu-driven program and is arranged with great flexibility so that the user can selectively move freely from one subroutine to another. The program 150 has a main routine indicated by block 151 which proceeds to a block 152 for storing data on the geometry of various types of fuel assemblies, and a block 153 for bringing up a main menu, which is illustrated in FIG. 20. When the main menu is called up, the program initially turns off all the "Soft Keys" or function keys, and then turns on those needed for the main menu selections. The menu is displayed and, on the actuation of any of the illustrated function keys to select a menu function, the keyborad is disabled, the time and date are printed out and the program proceeds to the selected subroutine. From the main menu, the user may select a number of functions, including a "Heading-Info" function at block 154, a "Print-Out" function at block 155, a "Copy-Data" function at block 156, an "Input-Data" function at block 157, an "Auto-Inspection" function at block 158 and a "Set-Up" function at block 159. Upon actuation of a selected "Soft Key", the main menu is turned off at block 160, and the program enters the selected subroutine. Initially, the user would typically call up the "Heading-Info" subroutine from the main menu, which subroutine is illustrated in FIG. 21. This subroutine turns off all the Soft Keys, turns on the Soft Keys needed for selection from a header menu and displays the header menu. From this menu, the user can select a "Plant-In" subroutine for entering plant identification information, a "Set-Time" subroutine for setting the current date and time and a "Get-FA-Type" subroutine for inputting information relative to the type of fuel assembly to be measured. Once that fuel assembly information is input, the program automatically proceeds to the block 152 to enter the geometry information for that type of fuel assembly. It will be noted that after each one of these menu functions is selected and the necessary information is input, the "0" Soft Key stores the input information and returns the user to the header menu. Each time a Soft Key is actuated from the main menu, the keyboard is disabled and the time is printed. The user can also turn off the main program loop from the "Heading-Infor" subroutine. In normal field operations, after the heading information has been entered, the "Set-Up" subroutine is operated to calibrate the underwater measurement assembly 50, including the LVDT gauges 135 thereof. Referring to FIGS. 1 and 27, the "Set-Up" subroutine, after performing reset functions, goes through a "Set-Return" function, turns off all Soft Keys and the knob functions, sets the keyboard interrupt to go to the "Set-Return" function and sets the error interrupt to go to an error recovery mode, which is self-explanatory. Then the subroutine turns on the Soft Keys needed for the "Set-Up" menu and displays the menu. The menu permits selection of a number of functions by the user. Typically, the user will first select the "Clamp-Nozzle" function, whereupon the program moves to block 161 (FIG. 19). This function activates a nozzle clamp solenoid valve for clamping the frame 80 to the fuel assembly top nozzle 33. The user may then perform a number of calibration functions. In a "Set-Carriage" function at block 162 (see FIG. 19) the operator, by keyboard control, drives the carriage assembly 120 to a number of known positions, preferably adjacent to the top, middle and bottom of the strongback channel 85, to calibrate the distance measuring apparatus with respect to the normal rest or zero position of the carriage 120 at the top of the frame 80. In the "Read-Standard" function at block 163, the user can check to see if the underwater measurement assembly 50 may have been distorted, during shipping, for example, The carriage 120 is moved to several predetermined checkpoints along a square standard (not shown) and the LVDT gauges 135 are read and, if the underwater measurement assembly 50 is out of square, appropriate compensations are made. In the "Set-LVDTS" function at block 164, each of the LVDT gauges 135 is engaged with the check plate 127 and readings are taken to calibrate the LVDT gauges 135. The carriage 120 is moved up to the top of the frame 80 and is then driven down to a predetermined location along the check plate. The LVDT gauges 135 are clamped on in their measuring positions, read, the reading is converted into inches and printed, and then the LVDT gauges 135 are released and the carriage 120 is moved down to the next position. In performing each of these calibration functions, the user can selectively drive the carriage 120 up and down by appropriate selections on the "Set-Up" menu (FIG. 27). For example, to move the carriage 120 up, Soft Key 5 would be actuated, whereupon the program would proceed to the "Move-Carriage" function at block 165 (FIG. 19) to drive the carriage 120 up until the user tells it to stop. The "Jog-Carriage" function can be selected to move the program to block 166 (FIG. 19) for fine or precise positioning of the carriage 120 at a selected location. After the set up procedures have been completed, the user returns the main menu and selects the "Auto-Inspection" subroutine. Referring to FIG. 25, the "Auto-Inspection" subroutine first turns off the Soft Keys and resets a BCD card and then checks to see if the control box is set in the automatic mode. If it is not, the program informs the operator and waits for a response and then again checks the control box. If it is in the automatic mode, the program sets the keyboard interrupt to go to point A and sets the error interrupt to go to a self-explanatory error recovery mode. The program sets fixed coordinates for predefined points on the underwater measurement assembly 50, then turns on Soft Keys necessary for the auto menu then displays that menu. Through the auto menu (FIG. 25), the user can select a "Clamp-Nozzle" function (block 167, FIG. 19), which is similar to that described above with respect to the "Set-Up" subroutine; an "ASM-In" function (block 168) for inutting fuel assembly identification information; a "Grid-In" function (block 169) for inputting information regarding the depth of the fuel assembly grids 35 below the top nozzle 33; and a "Get-Y-Corn" function (block 170) for inputting data on the location of a predetermined corner of the fuel assembly 30, thereby defining its rotational orientation. After the necessary information has been input, the user selects the "Auto-Control" subroutine, illustrated in FIG. 26. In the "Auto-Control" subroutine, the program automatically conducts an inspection of the fuel assembly 30, taking the necessary measurements for determining bow, twist, and tilt. The subroutine first turns off the Soft Keys, sets the error recovery, keyboard interrupt and a predetermined Soft Key to go to a "STPP" routine to stop the program until a user command is received. Next, the subroutine checks to see if the control box is set in the automatic mode. If it is not, it informs the operator and waits for a response. If it is, it clamps the LVDT gauges 135 to the top nozzle, reads them and releases them. With the data read the program finds and prints the centroid of the top nozzle and prints out that information along with the appropriate heading, then automatically moves the carriage 120 down to the next grid 35. At that grid, the clamping, reading, releasing, calculating and printing functions are repeated and then the program asks if all the grids have been completed. If they have not the program continues driving the carriage down to each grid until all the grids have been completed. Then the subroutine finds and prints the tilt and bow for each grid and drives the carriage back up to the top nozzle and stores all the results, and returns to the beginning of the "Auto-Inspection" subroutine and then to the auto menu. Throughout the "Set-Up" and "Auto-Inspection" subroutines, whenever the LVDT gauges are read, the readings are also converted from volts to inches by a "Convert-LVDT" function. The program 150 also makes it possible to manually key data into the computer, by selection of the "Input-Data" subroutine from the main menu. This might be necessary, for example, if the computer were not operating during a field inspection, in which case the measurements read would have to be written down and later keyed into the computer. Referring to FIG. 24, the "Input-Data" subroutine first sets the keyboard interrupt to go to a point A, sets the program to convert raw data to LVDT gauge position, turns on the necessary Soft Keys for the input menu and displays that menu. The user can select "Get-Y-Corn" and "Grid-In" functions which are the same as those described above in connection with "Auto-Inspection" subroutine. By selecting the "Data-In" function the user can key in the LVDT gauge data. The "Beta-In" function is for entry of a correction factor in the event that the strongback channel 85 is not perfectly vertical, while the "Cal-Factr-In" function is for entry of a calibration factor for the LVDT gauges if they are not standard. When all the necessary data and information has been keyed in, the user can direct the computer to find and print the centroids, bow, twist and tilt for the fuel assembly, after which the program will exit the menu and update the time on the screen. Alternatively, the user can direct the program to go to point B where it will immediately find and print the centroids, bow, twist and tilt and then return to the input menu. When it is desired to print out data independently of an "Auto-Inspection" function, the user can select the "Print-Out" subroutine from the main menu. Referring to FIG. 22, this subroutine turns off the Soft Keys, gets the record to be printed from the disk storage, converts the data and calculates and prints the centroids, bow, twist and tilt and then returns to the main menu. The "Copy-Data" subroutine (FIG. 23) may be selected from the main menu for the purpose of transferring data from one disk drive to another for providing backups at the end of each day's work. It is a significant aspect of the present invention that it accommodates fuel assembly types of different lengths and different envelope dimensions. Preferably, the system will accommodate fuel assembly envelope deviations of +/-2.00 inches from the vertical reference line 140, the measurements being accurate to within +/-0.030 inches. While, in the preferred embodiment, the underwater measurement assembly 50 is mounted from the pool deck 24 it could also be floor mounted at the bottom of the pool 23. The deck mounted configuration disclosed permits the entire underwater measurement assembly 50 to be adjusted and leveled from the deck 24 without underwater tools. It is another aspect of the invention that all measurements are taken with the fuel assembly 30 in an essentially free hanging condition on a handling crane. There is no need to release and re-engage the fuel assembly 30 with the overhead handling crane. Furthermore, the fuel assembly measurements can be made without a need to clamp the fuel assembly 30 in a precise location, since all measurements are made relative to the position of the top nozzle 33. Additionally, it will be noted that the system components are relatively lightweight and portable for service applications. The measurement system 40 utilizes a fully automatic computer control to keep test time under about five minutes per fuel assembly 30. This permits the system 40 to keep up with the refueling operations. An override control may also be included to allow a manual mode of operation utilizing the TV camera 126 for monitoring the measurement operation. From the foregoing, it can be seen that there has been provided an improved apparatus which permits determination of the bow, twist and tilt of a nuclear fuel assembly, the system being lightweight, accurate, easy to use and capable of manual or automatic remote operation, and being readily adaptable to fuel assembly types of different lengths and envelope dimensions, and which does not require disconnection of the fuel assembly from its handing tool or clamping of the fuel assembly in a precise location. |
claims | 1. An optical imaging device having an object plane and an image plane, the optical imaging device comprising:a housing having an interior and an exterior;a plurality of optical elements within an interior of the housing, the plurality of optical elements configured to image the object plane into the image plane via light passing along a beam path;a plurality of diaphragms, each diaphragm having an opening; anda diaphragm device in the exterior of the housing, the diaphragm device comprising a diaphragm store configured to hold the plurality of diaphragms arranged in a stack, each diaphragm being movable between a first position and a second position independently of the position of the other diaphragms,wherein for each diaphragm:in its first position, the diaphragm is in the diaphragm store, and the opening of the diaphragm is outside the beam path; andin its second position, the diaphragm is in the interior of the housing, and the opening of the diaphragm device is in the beam path. 2. The optical imaging device of claim 1, wherein:the plurality of diaphragms comprises a second diaphragm;the opening of the first diaphragm has a fixed geometry;the opening of the second diaphragm has a fixed geometry; andthe fixed geometry of the opening of the first diaphragm is different from the fixed geometry of the opening of the second diaphragm. 3. The optical imaging device of claim 2, wherein the optical imaging device is configured to be selectively stopped down based on the position of the first diaphragm and the position of the second diaphragm. 4. The optical imaging device of claim 1, wherein each diaphragm comprises a revolving disc diaphragm. 5. The optical imaging device of claim 4, wherein the diaphragm store comprises a plurality of separate plug-in units, and each revolving disc diaphragm is storable in a respective one of the individual plug-in units. 6. The optical imaging device of claim 4, wherein the housing includes an opening configured to allow diaphragms to be exchanged between the diaphragm store and the interior of the housing, and the revolving disc diaphragm stack is displaceable relative to the opening in the housing to position the first diaphragm in its second position. 7. The optical imaging device of claim 6, wherein, for each revolving disc diaphragm, the diaphragm device is configured to:remove the revolving disc diaphragm from its corresponding plug-in unit to introduce the revolving disc diaphragm into the beam path independently of the position of the other revolving disc diaphragms; andremove the revolving disc diaphragm from the beam path to position the revolving disc diaphragm into its corresponding plug-in unit independently of the position of the other revolving disc diaphragms. 8. The optical imaging device of claim 6, wherein, for each revolving disc diaphragm, the diaphragm device comprises a robot arm configured to:remove the revolving disc diaphragm from its corresponding plug-in unit to introduce the revolving disc diaphragm into the beam path independently of the position of the other revolving disc diaphragms; andremove the revolving disc diaphragm from the beam path to position the revolving disc diaphragm into its corresponding plug-in unit independently of the position of the other revolving disc diaphragms. 9. The optical imaging device of claim 8, wherein the diaphragm device comprises a lifting device configured to position a revolving disc diaphragm in the beam path, and the lifting device is configured to pick up the revolving disc diaphragm from the robot arm. 10. The optical imaging device of claim 9, wherein the lifting device is configured to move the revolving disc diaphragm via a rocking steering movement. 11. The optical imaging device of claim 9, wherein the lifting device defines a set of scales. 12. The optical imaging device of claim 9, wherein the lifting device defines a parallelogram guide. 13. The optical imaging device of claim 9, wherein the lifting device has a pantographic design. 14. The optical imaging device of claim 9, wherein the lifting device comprises solid joints. 15. The optical imaging device of claim 4, wherein the diaphragm device comprises a lifting device configured to position a revolving disc diaphragm in the beam path. 16. The optical imaging device of claim 4, wherein the diaphragm device comprises a holding device configured hold a revolving disc diaphragm in the beam path. 17. The optical imaging device of claim 4, wherein an optical element comprises a holding device configured to hold a revolving disc diaphragm in the beam path. 18. The optical imaging device of claim 4, wherein the diaphragm device comprises a lifting device which comprises a holding device configured to hold a revolving disc diaphragm in the beam path. 19. The optical imaging device of claim 18, further comprising spring elements configured to press the lifting device against the holding device to dynamically decouple a revolving disc diaphragm from the optical elements. 20. The optical imaging device of claim 18, wherein the revolving disc diaphragm is configured to be held by magnetic forces to dynamically decouple a revolving disc diaphragm from the optical elements. 21. The optical imaging device of claim 20, wherein the lifting device is dynamically decoupled from the optical elements. 22. The optical imaging device of claim 4, wherein the housing has an opening through which a revolving disc diaphragm is movable between the diaphragm store and the beam path. 23. The optical imaging device of claim 1, wherein the diaphragm device is dynamically decoupled from the optical elements. 24. The optical imaging device of claim 1, wherein the diaphragm store comprises a strip wound onto rollers. 25. The optical imaging device of claim 24, wherein the strip has a plurality of openings, and the strip is movable in the beam path to introduce the openings into the beam path by rotating the rollers. 26. The optical imaging device of claim 1, wherein the plurality of optical elements comprises a plurality of mirrors, and the light comprises EUV light. 27. A machine, comprising:an illuminating system; andan optical imaging system according to claim 1,wherein the illuminating system is configured to illuminate the object plane of the optical imaging system, and the machine is a projection exposure machine. 28. The machine of claim 27, wherein the plurality of optical elements of the optical imaging system comprises a plurality of mirrors, and the light comprises EUV light. 29. An optical imaging device having an object plane and an image plane, the optical imaging device comprising:a housing having an interior and an exterior;a plurality of mirrors within an interior of the housing, the plurality of mirrors configured to image the object plane into the image plane via EUV light passing along a beam path;a plurality of diaphragms, each diaphragm having an opening; anda diaphragm device in the exterior of the housing, the diaphragm device comprising a diaphragm store configured to hold the plurality of diaphragms arranged in a stack, each diaphragm being movable between a first position and a second position independently of the position of the other diaphragms,wherein for each diaphragm:in its first position, the diaphragm is in the diaphragm store, and the opening of the diaphragm is outside the beam path; andin its second position, the diaphragm is in the interior of the housing, and the opening of the diaphragm device is in the beam path; andthe diaphragm store comprises a plurality of separate plug-in units;each plug-in unit has a corresponding diaphragm; andfor each diaphragm, the diaphragm device is configured to:remove the diaphragm from its corresponding plug-in unit to introduce the diaphragm into the beam path independently of the position of the other diaphragms; andremove the diaphragm from the beam path to position the diaphragm into its corresponding plug-in unit independently of the position of the other diaphragms. 30. The optical imaging device of claim 29, wherein each diaphragm comprises a revolving disc diaphragm. |
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048470082 | description | DETAILED DESCRIPTION OF THE INVENTION All parts, percentages, ratios and proportions are on a weight basis unless otherwise stated herein or obvious herefrom to one ordinarily skilled in the art. Pure lead phosphate glass (i.e., a glass that does not contain any nuclear waste or iron) can be prepared by melting together PbO (lead oxide) and P.sub.2 O.sub.5 (phosphorus oxide) at elevated temperatures. The composition of the resulting glass can be varied by varying the ratio of the weight of PbO to the weight of P.sub.2 O.sub.5. However, if the weight percent of lead oxide exceeds about 66 weight percent a crystalline form of lead phosphate, not a glass, is formed. Hence, this composition (66 weight percent of PbO, and 34 weight percent of P.sub.2 O.sub.5) represents a critical limit in the sense that compositions which contain larger amounts of lead oxide no longer form a glass. The lower limit on the minimum amount of PbO which can be melted together with P.sub.2 O.sub.5 to form a suitable host glass for nuclear waste is important, although not as clear cut. The composition consisting of about 45 weight percent PbO and 55 weight percent of P.sub.2 O.sub.5 was taken to be a practical lower limit on the amount of lead oxide needed, since the viscosity of the molten glass became much larger as the PbO content was reduced further. The higher the viscosity, the harder the glass is to pour and the higher the required processing temperature becomes. The addition of simulated nuclear waste to the pure lead phosphate host glass does not substantially modify the lead oxide and phosphorous oxide limits discussed above. It was found, however, that the addition of simulated radioactive nuclear waste containing Fe.sub.2 O.sub.3 to the lead phosphate host glass produced a dramatic decrease in the corrosion rate. That is, pure lead phosphate glasses (i.e., with no Fe.sub.2 O.sub.3 containing nuclear waste added) are quite susceptable to aqueous corrosion. When the simulated radioactive nuclear waste containing Fe.sub.2 O.sub.3 was added to the lead phosphate host glass, however, a highly corrosion resistant and stable nuclear waste glass was formed. The lead phosphate glass appears to be insensitive to the details of the preparation procedure and can be made with a very large variation in the molar ratio of PbO to P.sub.2 O.sub.5 as indicated. For example, specimens of homogeneous lead-iron phosphate nuclear waste glasses loaded with 15 weight percent of simulated radioactive nuclear waste material have been prepared with the amount of PbO in the lead-iron phosphate host glass varied from 45 to 66 weight percent, the amount of P.sub.2 O.sub.5 varied from 34 to 55 weight percent, and the amount of Fe.sub.2 O.sub.3 varied from 0 to 10 weight percent depending on the iron content of the simulated nuclear waste. Most of the decrease in the corrosion rate of the lead-iron phosphate nuclear waste glass is due to the large amounts of iron oxide present in the radioactive nuclear waste material (for example, at one nuclear facility about 50 weight percent of the radioactive nuclear waste material is iron oxide Fe.sub.2 O.sub.3). The effects of various amounts of iron oxide on the corrosion rate of lead phosphate are shown in FIG. 3. As can be easily seen from FIG. 3, the addition of 9 weight percent Fe.sub.2 O.sub.3 to a pure lead phosphate glass improves the corrosion resistance by a factor of about 10,000. Hence, by purposely adding about 9 weight percent iron oxide to pure lead phosphate glass, one can produce a very stable and easily prepared glass, which can then be used to immobilize other types of radioactive nuclear waste material which do not contain large amounts of iron oxide. These wastes include reprocessed commercial nuclear power reactor wastes. FIG. 3 shows that the corrosion rate is substantially reduced by including at least about 9 weight percent of iron oxide (Fe.sub.2 O.sub.3) in the lead phosphate glass composition. This application has been successfully demonstrated in experiments where both simulated high-level defense nuclear waste and simulated radioactive nuclear-power reactor wastes were added to the lead-iron phosphate host glass. The resulting nuclear waste glass was a highly corrosion resistant and stable wasteform. The combining of radioactive nuclear waste with lead-iron phosphate glass forms a nuclear waste glass that is highly corrosion resistant, not susceptible to devitrification, and that can be prepared at a relatively low temperature. The presence of a high level of Fe.sub.2 O.sub.3 is critical. This type of synergistic effect, in which the corrosion resistance of the combined material is enhanced, also occurs in the case of borosilicate glass waste forms in that the waste loaded glass is significantly more stable than a glass formed from the pure borosilicate glass frit. In fact, the pure borosilicate host glass typically corrodes about 10 times faster than the glass in combination with simulated nuclear waste. [See: Sales, B.C., L.A. Boatner, H. Naramoto and C.W. White, J. Non-Cryst. Solids 53 (1982) 201; and Clark, D.E., C.A. Mauer, A. R. Jurgenson and L. Urwongse in Scientific bases for Nuclear Waste Management, Vol. 11, ed. W. Lutze (Elsevier North Holland, New York, 1982) pp. 1 to 14.] For the lead-iron phosphate waste form, however, the improvement in corrosion resistance following addition of the simulated iron-containing waste is much greater. Lead-iron phosphate glass is quite suitable as a long-term storage medium for high-level nuclear waste. The properties of lead-iron phosphate nuclear waste glass are superior to a borosilicate nuclear waste glass, which was recently selected for the long-term storage of some high level nuclear defense wastes. The borosilicate nuclear waste glass therefore is used herein as a standard to which new wasteform of the invention is compared. The invention provides a stable primary containment medium for disposal of high-level radioactive nuclear waste. The invention wasteform typically comprises a lead-iron phosphate glass containing up to 20 weight percent of nuclear waste of the type typically consisting of 50 weight percent of Fe.sub.2 O.sub.3, 9.8 weight percent of Al.sub.2 O.sub.3, 13.8 weight percent of MnO.sub.2, 4.5 weight percent of U.sub.3 O.sub.8, 3.7 weight percent of CaO, 6.2 weight percent of NiO, 1.2 weight percent of SiO.sub.2, 7.1 weight percent of Na.sub.2 O, 1 weight percent of Cs.sub.2 O, 1 weight percent of SrO and 1.3 weight percent of Na.sub.2 SO.sub.4 (or other nuclear waste mixtures with similar compositions). Such compositions with varying amounts of iron and aluminum represents a class of nuclear defense wastes. In addition, the lead-iron phosphate nuclear waste glass can typically be prepared containing 10 weight percent, of the above composition plus 5 weight percent of a composition that is representative of the waste generated by nuclear power reactors. In distilled water at 90.degree. C., the net release of all elements from both types of lead-iron phosphate nuclear waste glasses are 100 to 1000 times smaller (depending on the specific element) than the corresponding amounts released by a comparably loaded borosilicate glass wasteform. EXAMPLE Several lead-iron phosphate glasses were prepared incorporating either simulated radioactive defense nuclear waste or simulated reprocessed commercial waste combined with simulated radioactive defense waste to demonstrate the invention. Appropriate amounts of PbO and (NH.sub.4).sub.2 HPO.sub.4 powders were thoroughly mixed with 15 weight percent of a powdered form of a simulated metal oxide nuclear waste and melted in a platinum crucible at temperatures between 800.degree. and 1050.degree. C. for 3 hours. See Table II for the compositions. The compositions of the lead-iron phosphate host glass studied are given in Table I. The molten glass was then poured into a heated mold of spectroscopically pure carbon, annealed at 450.degree. C. for 2 hours and cooled to room temperature over the space of a few hours. All of the components of the waste were readily dissolved in a short time at 1050.degree. C., and all of the components except Al.sub.2 O.sub.3 and ZrO.sub.2 were dissolved at temperatures between 800.degree. and 900.degree. C. The lead-iron phosphate glass wasteforms prepared at 800.degree. to 900.degree. C. in which Al.sub.2 O.sub.3 and ZrO.sub.2 were not completely dissolved, however, were as corrosion resistant as the lead-iron phosphate wasteforms prepared at 1050.degree. C. All of the lead-iron phosphate glasses loaded with the simulated nuclear waste had a black appearance that resembled that of waste-loaded borosilicate glass. The lead-iron phosphate glasses that were heated to between 1000.degree. and 1050.degree. C. were very homogeneous. Corrosion tests of the type (MCC-1) developed by the Materials Characteristics Center located at Battelle Northwest Laboratories were used to compare the corrosion behavior of the lead-iron phosphate nuclear wasteform with that of an identically loaded borosilicate glass nuclear wasteform. Each wasteform was corroded for one month in distilled water at 90.degree. C. The results are shown in FIG. 1. The data shows that the net release of all of the elements from the lead-iron phosphate wasteform was at least 100 to 1000 times smaller than the corresponding amounts released by the borosilicate wasteform (that is, Frit 131 plus 29 percent of the first simulated nuclear waste composition--see Table II for the exact compositions). The concentrations of all of the elements present in the lead-iron phosphate leachate were below the detectable limits of the standard analytical chemical techniques employed (in this case, inductively coupled plasma emission analysis--ICP, atomic absorpotion, and flourimetry). The presence of iron (a component of the nuclear waste material) is primarily responsible for the very high corrosion resistance of the nuclear waste glass relative to that of pure lead phosphate glass. In more detail, FIG. 1 shows the 30-day corrosion rates at 90.degree. C. in distilled H.sub.2 O for lead-iron phosphate [Pb(PO.sub.3).sub.2 plus 15 weight percent of the first simulated nuclear waste] and borosilicate (Frit 131 plus 29 weight percent of the first simulated nuclear waste) nuclear waste glasses. The lead-iron phosphate and borosilicate nuclear waste glasses had the same waste per volume loading. The effects of the pH of the corroding solution on the corrosion rate of the lead-iron phosphate wasteform was also investigated (see FIG. 2) and compared to the behavior of a borosilicate glass wasteform. The lead-iron phosphate wasteform was comprised of 50 weight percent of PbO and 50 weight percent of P.sub.2 O.sub.5 plus 15 weight percent of the first simulated nuclear waste (see Table II). The waste weight percentages for the lead-iron phosphate glass versus borosilicate glass yield comparable waste per volume factors due to the higher density of the lead-iron phosphate glass (i.e., 5.+-.0.1 g/cm.sup.3) relative to borosilicate glass (2.6 g/cm.sup.3). The borosilicate glass was comprised of Frit 131 plus 9 weight percent of the first simulated nuclear waste (see Table II). In the neutral pH regions (i.e., for pH values between 5 and 9) which encompass the pH range of most natural ground waters, the corrosion rate of the lead-iron phosphate wasteform was 100 to 1000 times smaller than the corresponding corrosion rates of the borosilicate glass wasteform. At the pH extremes of 2 to 12, the corrosion rate of the lead-iron phosphate wasteform approaches but does not exceed that of the borosilicate glass wasteform (see FIG. 2). TABLE I ______________________________________ Lead-iron phosphate host glass compositions. The nuclear waste glass is formed by melting the lead-iron phosphate host glass together with a powdered form of the nuclear waste. Compound Weight % ______________________________________ PbO 40-66 P.sub.2 O.sub.5 30-55 Fe.sub.2 O.sub.3 *.sup.1 0-10 ______________________________________ Note: .sup.1 Amount of iron oxide added depends on type of highlevel nuclear waste. TABLE II __________________________________________________________________________ Lead-Iron Phospate And Borosilicate Nuclear Waste Glass Compositions Typical Lead-Iron Frit 131 Composition First Simulated Nuclear Second Simulated Nuclear Phosphate Compositions (Borosilicate Glass) Waste Composition Waste Composition (weight percent) (weight percent) (weight percent) (weight percent) __________________________________________________________________________ PbO 40-66 P.sub.2 O.sub.5 30-55 Fe.sub.2 O.sub.3 0-10 SiO.sub.2 57.9 Fe.sub.2 O.sub.3 50.0 ZrO.sub.2 12.10 B.sub.2 O.sub.3 14.7 Al.sub.2 O.sub.3 9.8 MoO.sub.3 12.67 Na.sub.2 O 17.7 MnO.sub.2 13.8 Nd.sub.2 O.sub.3 11.6 Li.sub.2 O 5.7 U.sub.3 O 4.5 CeO.sub.2 8.13 MgO 2.0 CaO.sup.8 3.7 RuO.sub.2 10.27 TiO.sub.6 1.0 NiO 6.3 Cs.sub.2 O 7.05 ZrO.sub.2 0.5 SiO.sub.2 1.3 U.sub.5 O.sub.8 5.54 La.sub.2 O.sub.3 0.5 Na.sub.2 O 7.2 La.sub.2 O.sub.3 3.6 Na.sub.2 SO.sub.4 1.3 Pr.sub.2 O.sub.3 3.6 Cs.sup.2 O 1.0 Sm.sub.2 O.sub.3 2.2 SrO.sup.2 1.0 Fe.sub.2 O.sub.3 3.7 P.sub.2 O 1.64 SrO.sup.5 2.59 BaO 3.83 PdO 3.65 TeO.sub.2 1.44 Y.sub.2 O.sub.3 1.46 Other Oxides 5.0 __________________________________________________________________________ Other tests indicated that the corrosion behavior of the lead-iron phosphate wasteform was not affected by large doses of gamma radiation, nor was the material unusually susceptible to corrosion at a higher temperature, e.g., 135.degree. C. In the 800.degree. to 1050.degree. C. temperature range, the viscosity of the molten lead-iron phosphate wasteform is much less than the prototype borosilicate glass wasteform, as evidenced by the fact that the lead-iron phosphate could be easily poured at 800.degree. C. In spite of the low viscosity for the lead-iron phosphate between 800.degree. to 1000.degree. C., the phosphate glass wasteform softened at 600.degree. C. which was about 25.degree. C. higher than the softening point of the borosilicate glass wasteform. A lead-iron phosphate wasteform was exposed to air at 550.degree. C. for 60 hours in order to determine if there was any rapid tendency to devitrify. No obvious devitrification of the wasteform was detected using X-ray diffraction analysis, and a subsequent corrosion test on the sample showed no degradation in corrosion resistance. A similar test of borosilicate glass treated at 500.degree. C. for 60 hours indicated that the borosilicate glass corrosion rate was measurably higher following the heat treatment. Since higher temperatures are needed to completely dissolve the Al.sub.2 O.sub.3 present in some radioactive nuclear wastes, an aluminum base alloy can be employed as a cannister material. A lead-iron phosphate glass wasteform of the invention was melted at 800.degree. C. in accordance with this invention and poured into a pure aluminum container (aluminum melts at 660.degree. C.). The aluminum container did not melt. The foregoing prescription of preferred embodiments of the invention has been presented for purposes of illustration and description. It is not intended to be exhaustive or to limit the invention to the precise form disclosed, and obviously many modifications and variations are possible in light of the above teachings. The embodiments were chosen and described in order to best explain the principles of the invention and its practical application to thereby enable others skilled in the art to best utilize the invention in various embodiments and with various modifications as are suited to the particular use contemplated. It is intended that the scope of the invention be defined by the claims appended hereto. |
abstract | A method and a system for detecting the presence of special nuclear materials in a suspect container. The system and its method include irradiating the suspect container with a beam of neutrons, so as to induce a thermal fission in a portion of the special nuclear materials, detecting the gamma rays that are emitted from the fission products formed by the thermal fission, to produce a detector signal, comparing the detector signal with a threshold value to form a comparison, and detecting the presence of the special nuclear materials using the comparison. |
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abstract | A filter device is for a collimator of an irradiation device. In an embodiment, the filter device includes a plurality of filter disks, arranged on a disk that is rotatable around an axis. Each filter disk is mounted so as to be movable in a radial direction and is tensioned by way of a respective spring element radially away from the axis or toward the axis against a guide contour, effecting a radial movement of at least one filter disk. |
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055457940 | abstract | Disclosed is a method for removing radioactive contaminants from metal surfaces by applying steam containing an inorganic acid and cerium IV. Cerium IV is applied to contaminated metal surfaces by introducing cerium IV in solution into a steam spray directed at contaminated metal surfaces. Cerium IV solution is converted to an essentially atomized or vapor phase by the steam. |
056087683 | description | BEST MODE FOR CARRYING OUT THE INVENTION With reference to FIG. 1, a conventional fuel assembly 10 includes a plurality of elongated, full length fuel rods (FLR's) 12 supported between a lower tie plate 14 and an upper tie plate 16. Fuel rods 12 pass through a plurality of fuel rod spacers 18 which provide intermediate support to retain the elongated rods in spaced relation and to restrain them from lateral vibration. Each fuel rod 12 comprises an elongated tube containing the fissile fuel (such as uranium or plutonium dioxide) in the form of pellets, particles, powder or the like, sealed in the tube by upper and lower end plugs 20 and 22. Lower end plugs 22 are formed with a taper for registration and support in support cavities 24 formed in the lower tie plate 14. Upper end plugs 20 are formed with extensions 26 which register with support cavities 28 in the upper tie plate 16. A channel 30 encloses the bundle in the usual manner. Several of the support cavities 24 in the lower tie plate 14 are formed with threads to receive tie fuel rods or PLR's (one shown at 32 in FIG. 1). As already mentioned hereinabove, there are typically eight tie rods and as many as fourteen PLR's in current fuel bundle designs which are attached to the lower tie plate by threaded holes and threaded end plug shanks. These rods typically terminate adjacent one of the spacers 18 as shown in FIG. 1. FIG. 2 illustrates a conventional end plug 34 of the type used with, for example, the PLR 32 shown in FIG. 1. The end plug 34, is formed of Zircaloy and includes a generally cylindrical upper body portion 36, including an annular fuel rod support shoulder 38 (the first rod is typically welded to the end plug), and a lower threaded portion 40 including a tapered lower end 42. This one piece end plug is adapted to be received in a threaded hole 44 formed in the lower tie plate 14. This conventional arrangement is illustrated in FIG. 3. It has been experienced that, after irradiation, many of the rods threaded into the lower tie plate 14 are difficult to remove. It has been determined that the cause of the binding or sticking of fuel rods within the lower tie plate is related to the corrosion which occurs on the male threads at the lower end 40 of the Zircaloy end plug 34. More specifically, the corrosion process creates a zirconium oxide with a net volume increase relative to the original volume of Zircaloy, and this excess material creates locking forces between the male end plug threads on the lower end 40 of the end plug, and the female threads in the hole 44 in the lower tie plate 14. Since the lower tie plate 14 is made of stainless steel (a harder and more corrosion resistant material than Zircaloy which does not experience any significant corrosion in the BWR environment), it has also been determined that the lower tie plate per se does not contribute to the locking problem. The invention here, and as best seen in FIG. 4, relates to a new Zircaloy end plug 46 which includes an upper body portion 48 which is similar to the upper body portion 36 of the conventional end plug 34, and a lower portion which includes a removable stainless steel connecting shank 50. The shank 50 is threaded at an upper section 52 and at a lower section 54 with a smooth intermediate portion 56 connecting the threaded sections. The upper threaded section 52 is receivable within a tapped hole 58 formed in the now axially shortened Zircaloy end plug upper body portion 48, and the lower threaded section 54 of the connecting shank 50 is adapted to be threaded directly into the correspondingly threaded hole 44 in the lower tie plate 14 as best seen in FIG. 5. In the preferred embodiment, the sections 52 and 54 of the stainless steel connecting shank 50 are threaded in opposite directions. At the same time, the tapped holes 44 and 58 in both the lower tie plate 14 and the Zircaloy end plug upper body portion 48, respectively, are correspondingly threaded. With this arrangement, when the fuel rod 32 (for example) is removed from the lower tie plate 14, both the end plug upper body portion 48 and end plug lower portion, i.e., the connecting shank 50, can be removed with the fuel rod 30, or the connecting shaft 50 can be left within the lower tie plate 14, depending on the direction of rotation of the fuel rod 32 relative to the end plug 46. In other words, if it is desired to maintain the stainless steel connecting shank 50 in the lower tie plate 14, then the fuel rod may be turned in, for example, a clockwise direction, thereby separating the fuel rod 30 and the end plug upper body portion 48 from the upper threaded section 52 of the shank 50, such that shank 50 remains in the lower tie plate 14. On the other hand, if it is desired to maintain the connecting shank 50 within the end plug upper body portion 48, then the fuel rod 30 is rotated in an opposite direction so that the fuel rod 32 and the entire end plug 46 including the upper connecting shank 50 are removed together from the lower tie plate 14. The above arrangement also provides an additional option in the event of any sticking problem which may be experienced when attempting to remove the fuel rod. In other words, if the threads on shank 50 are frozen in one direction of rotation of fuel rod 32, an option remains by virtue of being able to rotate the fuel rod in the opposite direction so that the fuel rod can be removed, with or without the shank 50. It is also a feature of this invention that if a concern exists relative to debris or dirt clogging an open threaded hole (e.g., hole 44) in the lower tie plate 14 during fuel bundle reconstitution work, then the stainless steel connector shank 50 can be left in the lower tie plate 14 rather than in the end plug body 48, thereby precluding dirt or debris from entering the lower tie plate 14. In the preferred embodiment, and as already noted above, the connecting shank 50 is preferably constructed of stainless steel but in any event, in order to minimize if not completely eliminate the corrosion and sticking problem discussed herein, the connecting shank material should be harder, stronger and more corrosion resistant than the Zircaloy end plug body material. Under certain conditions, it might be advantageous to favor one fuel rod removal technique over another and this can be facilitated by incorporating differential torque characteristics at opposite threaded ends of the connector shank 50. In addition, while the invention has been described above in connection with partial length fuel rods, it will be appreciated that the utilization of a connector shank 50 of the type disclosed herein may also be used with fuel bundle tie rods as well as full length fuel rods where appropriate. It will also be appreciated that the threaded end plugs 46 of this invention may be utilized not only with PLR's but with tie rods and FLR's as well. By utilizing threaded end plugs as disclosed herein for the FLR's, it will be appreciated that the biasing springs conventionally used to bias FLR's into the lower tie plate holes (utilizing non-threaded fuel rod end plugs) can be eliminated. Utilizing threaded end plugs to secure fuel rods within the lower tie plate also eliminates the need for conventional end plugs which protrude through the lower tie plate, and thus removes or eliminates concern for coolant flow impact resulting from such protruding end plugs. 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. |
053902173 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to carbon fiber-reinforced carbon composite materials, processes for their production and first walls of nuclear fusion reactors made of such carbon composite materials. 2. Discussion of the Background Carbon fiber-reinforced carbon composite materials (hereinafter referred to simply as C/C composite materials) are light in weight and highly strong and have a feature that they are excellent in the heat resistance and corrosion resistance. Therefore, they are used, for example, for aerospace materials such as rocket nozzles, nose cones or disk brakes of air planes, for heater elements, for hot pressing molds, for other mechanical parts, and for parts of nuclear reactors. Such C/C composite materials are usually prepared by impregnating or mixing a matrix material such as a thermosetting resin such as a phenol resin or a furan resin, or a thermoplastic resin such as pitch, to long or short carbon fibers of e.g. polyacrylonitrile or pitch type, followed by heating and molding, then baking the molded product in a non-oxidizing atmosphere such as an inert gas atmosphere at a temperature of from 600.degree. to 1,000.degree. C., and further densifying-the product by impregnating pitch or a resin thereto, followed by baking, or by a chemical vapor deposition method, or a combination of such methods, followed, if necessary, by graphitization. However, the resulting C/C composite materials were not necessarily satisfactory when they were used for the purposes of conducting or removing heat in one direction i.e. in the thickness direction, and they had problems in their practical application. For example, a first wall of a nuclear fusion reactor represents the entire structure in the nuclear fusion reactor, which is disposed to face the plasma, and includes e.g. limiters, diverters, blankets, and parts thereof. Such first wall is disposed close to the plasma and thus is under a severe environmental condition such that it is subjected to heat from plasma and bombernment of plasma particles. Particularly, the limiters and diverters receive high temperature loads, whereby the heat load conditions are particularly severe. As one of materials used for the first wall under such severe conditions, graphite may be mentioned. Graphite is an excellent low atomic number material from the viewpoint of plasma impurities and also has high thermal shock resistance. FIG. 4 in the drawings illustrates the most typical conventional first wall wherein graphite is used. In the illustrated first wall, a graphite tile 11 is secured to a metal substrate 3 by means of a fixing plate 8 and a connector 9. When heat from the plasma enters the graphite tile 11 facing the plasma, the heat is conducted to the substrate by the contact thermal conduction and also dissipated by thermal radiation. In such system, the graphite tile 11 and the substrate 3 are in contact with each other merely by the mechanical connection, and the thermal conductivity at the contact portion is not adequate, and cooling tends to be inadequate when the heat load is high or lasts for a long period of time. The conventional first wall has the following problems. When a high heat load (for example, 2 km/cm.sup.2 for 3 seconds, or 4 km/cm.sup.2 for at least one second) is exerted to the first wall, the surface temperature will be as high as at least about 2,800.degree. C. and the vapor pressure of the graphite tile will be at least about 10.sup.-3 atm, whereby the loss in thickness by sublimation from the surface of the graphite tile will be as large as about a few tends .mu.m/sec. As a result, inclusion of carbon atoms in the plasma increases, which brings about a problem that the control of plasma impurities will be thereby seriously adversely affected. Further, the loss of the graphite surface is substantial, whereby there is a problem that the useful life of the first wall is short. In conventional nuclear fusion reactors, it is rare that such high heat load is exerted to the first wall, and the conventional first wall may sufficiently provide its function against the above-mentioned problems. However, in order to further improve the level of safety, or for a future nuclear fusion reactor for which it is expected that a heat load higher than ever will be exerted to the first wall constantly over a long period of time, it is desired to develop a first wall having the above problems adequately solved. Under these circumstances, the present inventors have conducted various studies to overcome the above-mentioned drawbacks and to obtain a C/C composite material useful for the above-mentioned first wall or the like, and have finally arrived at the present invention. SUMMARY OF THE INVENTION The present invention provides: 1. A carbon fiber-reinforced carbon composite material, wherein carbon fibers are oriented substantially in the thickness direction, the ratio of the thermal conductivity in the thickness direction to the thermal conductivity in a direction perpendicular to the thickness direction is at least 2, and the thermal conductivity in the thickness direction is at least 3 W/cm.multidot..degree.C. 2. A process for producing a carbon fiber-reinforced carbon composite material, which comprises impregnating long carbon fibers with a thermosetting resin, followed by heating to obtain a fiber/resin composite, cutting the fiber/resin composite into pieces having a length longer than the thickness of the desired composite material, aligning the composite pieces in one direction substantially in parallel to one another, exerting a pressure to the aligned composite pieces in a direction perpendicular to the longitudinal direction of the fibers, molding them to cure the resin, followed by carbonization, and impregnating the carbonized product with pitch or with a thermosetting resin, followed by carbonization and, if necessary, by graphitization. 3. A carbon fiber-reinforced carbon composite material having a metal bonded thereto, wherein: (i) the carbon composite material comprises carbon fibers which are oriented substantially in the thickness direction, wherein the ratio of the thermal conductivity in the thickness direction to the thermal conductivity in a direction perpendicular to the thickness direction is at least 2, and the thermal conductivity in the thickness direction is at least 3 W/cm.multidot..degree.C., and PA1 (ii) the metal is bonded to one side of the carbon composite material which is substantially perpendicular to the thickness direction of the carbon composite material. 4. A method for producing a carbon fiber-reinforced carbon composite material having a metal bonded thereto, which comprises impregnating long carbon fibers with a thermosetting resin, followed by heating to obtain a fiber-resin composite, cutting the fiber/resin composite into pieces having a length longer than the thickness of the desired composite material, aligning the composite pieces in one direction substantially in parallel to one another, exerting a pressure to the aligned composite pieces in a direction perpendicular to the longitudinal direction of the fibers, molding them to cure the resin, followed by carbonization, impregnating the carbonized product with pitch or with a thermosetting resin, followed by carbonization and, if necessary, by graphitization, to obtain a carbon fiber-reinforced carbon composite material, and then bonding a metal to one side of the carbon composite material which is substantially perpendicular to the thickness direction of the carbon composite material. 5. A first wall of a nuclear fusion reactor to be disposed to face a plasma of the nuclear fusion reactor, which is composed essentially of a carbon fiber-reinforced carbon composite material wherein carbon fibers are oriented substantially in the thickness direction, the ratio of the thermal conductivity in the thickness direction to the thermal conductivity in a direction perpendicular to the thickness direction is at least 2, and the thermal conductivity in the thickness direction is at least 3 W/cm.multidot..degree.C., and which is to be disposed so that one side which is substantially perpendicular to the thickness direction, faces the plasma. 6. A carbon fiber-reinforced carbon composite material, wherein at least about 50% of carbon fibers are oriented substantially in the thickness direction, the ratio of the thermal conductivity in the thickness direction to the thermal conductivity in a direction perpendicular to the thickness direction and to the plane of orientation of the fibers in the thickness direction is at least 1.2, and the thermal conductivity in the thickness direction is at least 1.5 W/cm.multidot..degree.C. 7. A process for producing a carbon fiber-reinforced carbon composite material, which comprises impregnating a woven fabric of carbon fiber with a thermosetting resin to obtain a fiber/resin composite, cutting the fiber/resin composite into pieces having a desired size, overlaying the composite pieces one on another so that the plane of each woven fabric is in the thickness direction, followed by molding, curing and carbonization, and impregnating the carbonized product with pitch or with a thermosetting resin, followed by carbonization and, if necessary, graphitization. 8. A process for producing a carbon fiber-reinforced carbon composite material, which comprises overlaying woven fabrics of carbon fiber impregnated with a thermosetting resin one on another, followed by molding, curing and carbonization, and impregnating the carbonized product with pitch or with a thermosetting resin, followed by carbonization and, if necessary, graphitization to obtain a carbon fiber-reinforced carbon. composite material, cutting the carbon composite material into pieces having a desired size, overlaying the pieces one on another so that the plane of each woven fabric is in the thickness direction and binding or bonding them integrally. 9. A process for producing a carbon fiber-reinforced carbon composite material, which comprises fibrillating short carbon fibers to form webs, overlaying webs one on another so that the plane of each web is in the thickness direction, impregnating the webs with pitch or with a thermosetting resin before or after the overlaying, followed by molding, curing and carbonization, and impregnating the carbonized product with pitch or with a thermosetting resin, followed by carbonization and, if necessary, graphitization. 10. A process for producing a carbon fiber-reinforced carbon composite material, which comprises fibrillating short carbon fibers to form webs, needling the webs in the thickness direction to obtain felts, overlaying the felts one on another so that the plane of each felt is in the thickness direction of the desired carbon composite material, impregnating the felts with pitch or with thermosetting resin before or after the overlaying, followed by molding, curing and carbonization, and impregnating the carbonized product with pitch or with a thermosetting resin, followed by carbonization and, if necessary, graphitization. 11. A carbon fiber-reinforced carbon composite material obtained by the method of the above 9 or 10, wherein the thermal conductivity in the thickness direction and the thermal conductivity in a direction perpendicular to the thickness direction and parallel to the plane of each web or felt, are at least 3 W/cm.multidot..degree.C. 12. A first wall of a nuclear fusion reactor to be disposed to face a plasma of the nuclear fusion reactor, which is composed essentially of a carbon fiber-reinforced carbon composite material wherein at least 50% of carbon fibers are oriented substantially in the thickness direction, the ratio of the thermal conductivity in the thickness direction to the thermal conductivity in a direction perpendicular to the thickness direction and to the plane of orientation of the fibers in the thickness direction, is at least 1.2, and the thermal conductivity in the thickness direction is at least 1.5 W/cm.multidot..degree.C., and which is to be disposed so that one side which is substantially perpendicular to the thickness direction, faces the plasma. 13. A first wall of a nuclear fusion reactor to be disposed to face a plasma of the nuclear fusion reactor, which is composed essentially of a carbon fiber-reinforced carbon composite material obtained by the method of the above 9 or 10, wherein the thermal conductivity in the thickness direction and the thermal conductivity of a direction perpendicular to the thickness direction and parallel to the plane of each web or felt, are at least 3 W/cm.multidot..degree.C., and which is to be disposed so that one side which is substantially perpendicular to the thickness direction, faces the plasma. |
description | 1. Field of the Invention This invention is concerned with a fibre-reinforced plastic (FRP) honeycomb sandwich structure in which a honeycomb core is arranged parallel to panel surface in order to solve the peel-off problem between hull and core in the prior art and is concerned with a honeycomb core assembly set tank in which a lot of FRP internal tanks are gathered into honeycomb core assembly. 2. Description of the Prior Art As a social background, global warming and dryness of fossil resource are given. One answer with possibility to these large questions is a fuel cell power generation system. Everybody knows even if the highest technology of the modern civilization is made good use of, the fossil fuel that is the inheritance of earth resource cannot be reproduced. When we enjoy the modern civilization in every day living, it is inevitable and irreversible to consume the fossil fuel. The fuel cell generates electric energy from “Hydrogen (H2)” and “Oxygen (O2)”, and the exhaust is “Water (H2O)”. In the power generation process, the fuel cell system does not exhaust the carbon dioxide assumed to be a cause of global warming. Present days, “Hydrogen (H2)” used for the fuel cell is refined from LNG which is the fossil fuel. However, predecessors of the modern civilization left us the reproducing scientific theories for “Hydrogen (H2)” which is the fuel of the fuel cell. Even if the highest technology of the modern civilization is made good use of, the fossil fuel that is the earth resource cannot be reproduced. On the other hand, the power generation system of the fuel cell can be reproduced, even if it is now only the theory on the desk. Near future, the fuel cell has a possibility of being the energy system for our life to be an ideal energy system. The fuel cell is an ideal energy system, indeed. However, it is necessary to overcome the two big difficulties for the practical use. One is a power generation cell of the fuel cell, and another one is a hydrogen tank. The power generation cell of the fuel cell has been improved by marketing now. However, the development of the hydrogen tank is still difficult. Hydrogen is a gas at the normal temperature, and the liquefaction temperature in the atmospheric pressure is an extremely low temperature at degree of −200° C. or less. Moreover, the gas hydrogen of 4% is mixed with the oxygen in the air, and when the spark flies to neighborhood, it burns explosively. Handling is very difficult. Safety demands that the design and characteristics of the hydrogen tank be rated for world class performance. The hydrogen tank for the car has been researched in each car manufacturer, and the structural strength has achieved 750 atmospheric pressures now. However, there are some defects in the tank volume, shape and the spacing. The difficult problem when we put the fuel cell to practical use in our life is transportation of hydrogen. As for hydrogen, if it is not liquefied, the transportation efficiency is too inferior to liquid gasoline because molecular weight is very small. However, the liquid hydrogen is not obtained if it does not cool to the extremely low temperature of −200° C. or less. It costs much to transport the liquid hydrogen with the vehicle only for special use. Additionally, the difficult one is a hydrogen station. A similar hydrogen station to the gas station is necessary for the spread of the fuel cell powered vehicle. The internal pressure of the hydrogen tank in a state-of-the-art vehicle is 750 atmospheric pressures now. Therefore as for assumed hydrogen station, it is necessary to equip an underground tank with fully cooling system keeping the liquid hydrogen at minus 200° C. or less and also it is necessary to equip a pressurizing system which pressurizes the hydrogen gas to 750 atmospheric pressures. It is forecast that not only the installation cost but also maintenance cost becomes huge for the assumed hydrogen station. Moreover, it is difficult to confine the liquid hydrogen in an underground tank of the normal temperature when the cooling function is lost by any chance. As a result, the situation in which most of the liquid hydrogen preserved at the hydrogen station should be discharged in air is thought, too. Because the gas hydrogen is an explosiveness gas, it is large issue for safety. These are the problems and the current state that are demanded of the storing system for hydrogen gas. The fuel cell power generation system has the possibility of becoming the energy system that can be called ideal. However, the achievement of ideal energy system demands to us the research and development of cooperation and endurance. The honeycomb structure has a good deal of benefits in lightweight and rigidity, thus the honeycomb sandwich panel has been widely adopted as a structure material in the architectural field and the aircraft field. However, in a usual prior art the honeycomb core, a set of a lot of honeycomb cells, is manufactured from cardboard, mold plastic and lamina aluminum. Their structural strength are insufficient to support the vertical load on side surface, so, the honeycomb core cannot be arranged parallel to the surface in the honeycomb sandwich panel. In the prior art the direction of honeycomb core is arranged vertical to the panel surface, thus the hull is bonded on a small area of the six-corner edges of honeycomb core assembly. In a usual prior art, honeycomb sandwich panel cannot avoid the peel-off problem between the honeycomb core and hull. The peel-off problem is the fatal defect of the honeycomb sandwich panel. The patent No. 4862975 in Japan, a manufacturing method of honeycomb core made from FRP prepreg is indicated. And it shows the manufacturing process in which the honeycomb core assembly is arranged parallel to the surface of honeycomb sandwich panel. However, the manufacturing process of patent No. 4862975 is very complex and is inferior to productivity. And the shapes of six-corner honeycomb cells are easily distorted. This invention is directed to the application of a honeycomb core structure in the storing and transportation of sensitive and/or volatile materials, such as hydrogen gas in a hydrogen fuel system. In order to use a honeycomb structure, the honeycomb core of a honeycomb sandwich panel from a vertical direction to a horizontal direction to solve the peel-off problem in the prior art. The peel-off problem is caused from the vertical arrangement of honeycomb core assembly. It is absolutely necessary to arrange the honeycomb core in parallel to the surface of the honeycomb sandwich panel to solve the problem. However, when the honeycomb core is arranged parallel to the panel surface, the panel load becomes to hang on the side surface of honeycomb core assembly. As for the honeycomb core assembly in the prior art, strength is insufficient to support the load on the side surface of the honeycomb core assembly because it is manufactured from aluminum lamina, mold plastic and cardboards. It is necessary to manufacture the strong honeycomb core that will not be collapsed by the side load. In this invention, honeycomb core assembly is made from soft FRP prepreg (pre-impregnated fibre-reinforced plastic) with double structural wall and is manufactured by the processes of stiffening by heat and pressure. The pressure is generated from the reaction force between the heat expansion pressure by internal pressurizing devices and external frame structure that restricts mechanically all the surface of honeycomb core assembly. The new process of the invention also gives rise to a new usage where a honeycomb core assembly set tank may be manufactured that has many internal tanks in every honeycomb cell. This set tank is derived from the double wall honeycomb cells structure of this invention. The internal wall maintains the pressure of the internal tank, and the external wall of the double wall tank endures the external shock loading. Patent No. 4862975 in Japan shows an epoch-making manufacturing method of the honeycomb core made from FRP prepreg. And it also shows the manufacturing processes of arranging the honeycomb core assembly in parallel to the surface of honeycomb sandwich panel. However, the manufacturing process is complex and is inferior to productivity and the shapes of six-corner honeycomb cells are easily distorted. The honeycomb core material and the manufacturing processes shown in Patent No. 4862975 are as follows. FIG. 17 illustrates the manufacturing of the honeycomb core mother material shown in Patent No. 4862975. The stapler needles are continuously driven to sew between upper FRP prepreg and lower FRP prepreg, such that the stapler manufactures a continuous body of a long and slender bag. FIG. 18 shows the honeycomb core mother in which many cylindrical heat foam plastic resins are inserted, sequentially. It is also possible to insert an air tube. FIG. 19 shows the honeycomb structure in which the honeycomb core is horizontally arranged. The honeycomb core is manufactured by the method of Patent No. 4862975. As shown in the above-discussed illustrations of manufacturing using the teachings of Patent No. 4862975, it is complex to manufacture many long and slender bags repeatedly. Moreover, it is not productive to insert an air tube and heat expansion resin into a long and slender space. The insertion work of the heat foam plastic resin is easier than that of the air tube. In this case, the heat expansion resin remains in the honeycomb core after heat foam is processed. Therefore, the weight of honeycomb structure becomes heavier compared with the method of pressurizing by the air tube. The present invention is intended to solve these defects. For answering the problem and the current state demanded to the storing system of hydrogen gas, the processing technology and the concept of new hydrogen tank are described. In terms of the honeycomb structure itself, the most important problem that should be solved is the peel-off problem in the prior art. To solve the problem, it is necessary to arrange the honeycomb core horizontally to the panel surface of honeycomb sandwich panel. It is necessary to manufacture a lot of honeycomb cells that will not be collapsed by the vertical load, because the honeycomb core is a set of many honeycomb cells in assembly. The honeycomb cell that can withstand not collapsing by a vertical load cannot be manufactured in the prior art. Patent No. 4862975 indicated the method to manufacture the honeycomb cells by FRP prepreg. However, its manufacturing process is complex and is inferior in productivity. Thus, it was necessary to invent a new process of manufacturing. The invention is a manufacturing process for a FRP honeycomb structure in which a honeycomb core assembly is arranged parallel to the surface, for solving the peel-off problem of the honeycomb sandwich panel in the prior art. The peel-off problem is caused from the vertical arrangement of a honeycomb core assembly, because, when a honeycomb core assembly is vertically arranged, the hull of the honeycomb sandwich panel is bonded on a small area of the six-corner edges of honeycomb core assembly. Therefore, it is necessary to arrange the honeycomb core assembly in parallel to the surface of the honeycomb sandwich panel, for solving the peel-off problem. However, the panel load becomes to hang on the side surface of honeycomb core assembly, when honeycomb core assembly is arranged parallel to the panel surface. As for the honeycomb core assembly in the prior art, strength is insufficient to support the load on the side surface of the honeycomb core assembly because it is manufactured from aluminum lamina, mold plastic and cardboards. Therefore, it is necessary to manufacture the honeycomb core that will not be collapsed by the side load. In this invention, the honeycomb core assembly is made from soft FRP prepreg with a double structural wall and is manufactured by processes of stiffening by heat and pressure. And the pressure is generated from the reaction force between the heat expansion pressure by internal pressurizing devices and external frame structure that restricts mechanically all the surface of honeycomb core assembly. The soft FRP prepreg is a FRP prepreg material not stiffened yet. The manufacturing process of the FRP honeycomb structure is composed of two manufacturing processes. The first is the process for manufacturing soft FRP honeycomb structure by the soft FRP prepreg and the second is the process for stiffening the soft FRP honeycomb structure to a rigid FRP honeycomb structure by heat and pressure. The second invention is the manufacturing process for the honeycomb core assembly set tank in which a plurality of internal tanks are gathered in a honeycomb core assembly structure, for the purpose of manufacturing a tank with a double structural wall whose total capacity is considerably large. This set tank is derived from the manufacturing process for FRP honeycomb structures in which a honeycomb core assembly is arranged parallel to the surface. The internal wall maintains the pressure of the internal tank, and the outside wall of the every double wall tank endures the external shock loading. Theoretically speaking, the honeycomb core structure of six-corner cell can be infinitely arranged, and its structural position is unique. In a similar fashion, theoretically a set tank in which a lot of internal tanks are gathered into a honeycomb core assembly can be arranged infinitely, thus its total capacity can be arbitrarily large. Generally, there is a limitation in structural strength of the wall, wherein the diameter of the internal tank becomes large in low pressure and becomes small in high pressure, so the diameter of internal tank is varied in inverse proportion to the pressure of internal tank. The diameter of tank can be from 100 mm to 1000 mm roughly, in this invention. Because the longitudinal load is multiplication of the area of base and the internal pressure, the longitudinal load is unrelated to the tank length. Therefore, if the diameter of the tank endures internal pressure, theoretically there is little limitation in the length of internal tank. In this invention, the honeycomb core assembly set tank is made from soft FRP prepreg and is manufactured by processes of stiffening by heat and pressure. The pressure is generated from the reaction force between the heat expansion pressure by internal pressurizing devices and external frame structure that restricts mechanically all the surface of the honeycomb core assembly set tank. The soft FRP prepreg is a FRP prepreg material not stiffened yet. The manufacturing process of the honeycomb core assembly set tank is composed of two manufacturing processes. The first is the process for manufacturing soft honeycomb core assembly set tank by soft FRP prepreg and the second is the process for stiffening process from the soft honeycomb core assembly set tank to rigid honeycomb core assembly set tank by heat and pressure. In the prior art, an autoclave, a device that does heating and pressurizing at the same time, is used for the heating and stiffening procedure for FRP structure. The autoclave uniformly pressurizes the external surfaces of FRP structure by air pressure. The air pressure is provided from an outer supplier or is supplied by evaporation pressure inside the autoclave and it fills the whole internal space of the autoclave. The autoclave is suitable for manufacturing a high-pressure tank of a comparatively large diameter single wall tank. However, there is a limitation in the structural strength; the diameter of a high-pressure tank so constructed cannot be infinitely enlarged. Further, it is difficult to manufacture a set tank using the autoclave. Specifically, when a plurality of tanks are gathered in an assembly structure, there is a middle-stuffing object between the internal FRP wall and the external FRP wall. Therefore, there is a theoretical limitation in the capacity of the high-pressure tank manufactured by the autoclave. Also, a mass high-pressure tank is inferior in safety in the event of an accident. For example, the hydrogen tank of the vehicle powered by fuel battery of 70M Pascal manufactured by a Japanese automaker is a cocoon shaped tank with a single wall. The diameter is about 400 mm and the capacity is about 150 liters or less. The single tank lacks the extendibility. It is difficult to manufacture a FRP tank having double structural walls using the autoclave. The reason is as follows: As shown in FIG. 14, when a FRP tank is manufactured by an autoclave, the external surface of the tank is pressurized by air pressure in the autoclave. The internal surface of the tank is pressurized by the vapor pressure enclosed in the tank. The external surface of the FRP tank is manufactured by long and slender zonal soft FRP prepreg. The wall of tank is strongly pressed by the pressure between the internal vapor pressure and the external air pressure. The full strength of tank wall cannot be obtained when either of these pressures is lacking. Thus, the tank wall by the autoclave is a single structural wall. Referring to FIG. 15, when a FRP tank with a double structural wall is manufactured by the autoclave, the internal tank wall is pressurized only by the expansion pressure of internal tank and external wall is pressurized only by the air pressure in the autoclave. Between the internal FRP wall and the external FRP wall, there is a mid-stuffing object in which an internal pressure is not generated. The mid-stuffing object does not voluntarily generate any pressure. Therefore, the internal tank FRP wall cannot be pressed uniformly. Here also it is difficult to secure the strength of the tank wall. This invention is basically different from the autoclave that uniformly pressurizes the external surface of FRP structure. This invention adopts a new method as shown for example in FIG. 16 with the following features: (1) FRP honeycomb core is a set of the assembly of many honeycomb cells. (2) The base material of the individual honeycomb cell is the heat foam resin. (3) The honeycomb cell is manufactured by reinforcing the base material with soft FRP prepreg. (4) The FRP honeycomb core assembly is assembled with a lot of the honeycomb cells. (5) All surfaces of the honeycomb core assembly are restrained with an external frame. (6) The external frame is put in the heating oven and the external frame and honeycomb core assembly are heated to stiffening temperature. There is no air pressure device outside the external frame. (7) The heat foam resin expands by heating. (8) The FRP prepreg is stiffened by the heat expansion pressure and the reaction force of external frame. By this method, there is no middle part where internal pressure is not generated. Therefore, it becomes possible to manufacture the double wall structure of the internal wall and the external wall. The internal wall maintains the pressure of the internal tank. The external wall endures the external shock loading. The tank shape air pressure device assembly is manufactured from heatproof plastic material. It is expanded by internal pressure and temperature. The shape of the internal tank is transformed permanently. Moreover, an external wall is an adhesive where it has strength. The structure of theoretical infinity can be manufactured by six-corner element's tying like the honeycomb core, if an individual element has a six-corner shape. FIG. 16 uses the references as follows to illustrate the components of the invention: (BB) six-corner honeycomb cell assembly, (BC) five-corner honeycomb cell assembly, (DD) trapezoid filler assembly, (MCA) External frame assembly which restricts all the surface of honeycomb structure manufactured from soft FRP prepreg, (PIS) Internal FRP wall manufactured from soft FRP prepreg, (POS) External FRP wall manufactured from soft FRP prepreg, and (TAA) tank shape air pressure device assembly. The FRP honeycomb structure can be used for the structural material of a vehicle in which lightweight and high strength are demanded, such as an aircraft, a rapid-transit railway car and an automobile. As for the glass fiber FRP, raw material exists unlimitedly and its manufacturing facility does not cost much. FRP material also does not rust when exposed to seawater. Therefore, it is the best material for use in a large ship or for the wind power generation near the sea. Generally speaking, there is a limitation in structural strength of tank wall, the diameter of internal tank becomes large in low pressure and becomes small in high pressure. It is natural that a high-pressure tank becomes a set tank composed of a plurality of small diameter tanks. Also naturally, it becomes inevitable to improve the accumulation rate, for manufacturing the efficient high-pressure set tank. In this invention, there is no limitation on the number of internal tanks that will form a set tank having the honeycomb core assembly. Thus, the invention allows the manufacture of a high-pressure set tank whose total capacity is theoretically unrestricted. Further, with a high-pressure tank, it is necessary to provide against external shock loading at the same time as maintaining internal pressure. In this invention, an internal FRP wall maintains the pressure of the internal tank, and an external FRP wall endures the external shock loading. As for the set tank of this invention, the role of all structural materials is clear, and the external shock loading is allotted to all the structural materials in the set tank. In addition, a set tank of the small diameter tanks can divide the internal energy of an individual high-pressure tank. Therefore, the set tank of small diameter tanks is much safer than a single tank having a big diameter. It is useful for the high-pressure tank such as liquefied natural gases and the hydrogen gases. The manufacturing process in this invention can manufacture a small diameter single tank having an internal volume of 500 cc. A hydrostatic test was conducted on a single tank pressurized at 20 MPa. The connection part of the tank was destroyed in the hydrostatic test, but main body wall of the single tank was not destroyed. Similarly, a small diameter single tank with an internal volume of 900 cc of the content was manufactured using as the material glass fiber FRP prepreg. The specifications for the test piece were 175 mm in length, compression areas 104 mm×104 mm. The weight was 0.89 Kg. The diameter of the weight-reducing hole was φ77 mm. The maximum examination load of the uni-axial compression test was 295.5 kN. A straight-line part in the tank was cut for the strength test, and the compression examination was done. The test piece was not destroyed, though the test load reached the maximum value of the compression examination device. When applied to the storage and/or transportation of sensitive or volatile materials, such as in a hydrogen fuel system, at the room temperature, a weight density which is almost equal to the liquid hydrogen cooled to minus 200° C. can be obtained by pressurizing the gas hydrogen to about 750 atmospheric pressures. Equipment and technology for a hydrogen tank rated to 750 atmospheric pressures already exists in the automotive industry. Generally speaking, there is a limitation in the structural strength of tank wall material; the diameter of internal tank becomes large in low pressure and becomes small in high pressure. Thus the diameter of high-pressure tank is varied in inverse proportion to the pressure of internal tank. Roughly, the diameter of high-pressure tank is from 100 mm to 1000 mm. The capacity of a single tank is limited by the strength of material that composes the tank wall. For purposes of a hydrogen fuel system, a set tank may be used that is manufactured by combining a plurality of comparatively small diameter tanks. At a hydrogen station and on a lorry, a cooling system is not necessary if the transportation container has the ability to keep the hydrogen gas at 750 atmospheric pressures in the room temperature. Theoretically, the design load of tank wall doubles when the diameter of a high-pressure tank doubles when internal pressure is the same. Therefore, to manufacture a massive high-pressure tank the material of high strength, for instance, carbon fiber FRP is necessary. However, carbon fiber FRP is very expensive. According to the present invention, a set tank is assembled with many internal tanks, wherein the diameter of the internal tank is smaller than that of a big diameter high-pressure tank developed in automaker. Each internal tank is composed of internal wall and external wall. The external shape of the assembled set tank is a rectangular hexahedron. Theoretically, the design load of the tank wall becomes half when the diameter of a high-pressure tank becomes half when internal pressure is the same. Thus, the material of high strength is not necessary to manufacture a small diameter tank. It is enough in glass fiber FRP. The price of glass fiber FRP is about 1/10 compared with carbon fiber FRP. The design load of the tank wall is smaller than that of a big diameter high-pressure tank. The weight of individual tank wall is light. However, a set tank is an assembly of many small tanks, so the total weight becomes the same or a little heavy compared with the big diameter tank. A set tank according to the invention is composed of a plurality of internal tanks arranged as a honeycomb core assembly. Theoretically, the honeycomb core assembly of six-corners can be infinitely arranged. Therefore, theoretical total capacity of the set tank is infinity. Safety increases compared with the big diameter tank, because a set tank of many small tanks can divide the explosion energy to the sum of small energy when any trouble occurs. The automatic rolling device of the hydrogen tank has been completed, too. An automatic rolling of the tank can be done in ten minutes. The hydrogen tank is a cocoon type tank made of carbon fiber FRP. It has the resisting pressure of 750 atmospheric pressures. The capacity is 40 liters per one tank. Even 150 liters have been completed now. The hydrogen tank is one of the key technologies in accomplishing a hydrogen fuel cell power generation system. The present invention is directed to solving at least some of the problems associated with developing and implementing such a system cost effectively. Referring to the drawing as follows, it explains the form of concrete execution of the manufacturing process of the FRP honeycomb structure in which a honeycomb core of double wall is arranged parallel to the surface of honeycomb sandwich panel, and explains the form of execution of a honeycomb core assembly set tank structure in which a plurality of internal tanks of double wall are gathered. FIG. 1 shows a cylindrical air-type pressure device. The cylindrical air-type pressure device (1) is made from heatproof plastic tube and heatproof rubber tube; it has enough length and it encloses the heat evaporation compound (2). The heat evaporation compound (2) is the heat blowing agent and the evaporating liquids. The evaporating liquids are water and alcohol. Both ends of the cylindrical air-type pressure device (1) are sealed. Because the air-type pressure device (1) can be made from the tube of heatproof plastic material and heatproof rubber, the length of the air-type pressure device (1) is arbitrary. FIG. 2 shows a cylindrical air-type pressure device assembly. The cylindrical air-type pressure device assembly (4) is manufactured by wrapping the external surface of cylindrical air-type-pressure device (3) with a soft FRP prepreg (5) two or more times. The soft FRP prepreg (5) becomes the internal FRP wall of honeycomb core of double wall. At room temperature, the soft FRP prepreg (5) is the soft cloth, so it is not difficult to wrap the air-type pressure device (3) with the soft FRP prepreg (5). Because the adhesive of prepreg deteriorates at the room temperature, it is preferable to preserve the product within the freezer at minus 5° C. or less. FIG. 3 shows a solid-type pressure device. The solid-type pressure device is made from the heat foam plastic resin by the metal mold of pushing out. It has enough length (9) and has semicircle vacant space (10) inside it. The solid-type pressure device has three types of mold heat foam plastic resin. The first is six-corner solid-type pressure device of half cut (6) and the second is four-corner solid-type pressure device of half cut (7) and the third is trapezoid-corner solid-type pressure device (8). Because the solid-type pressure device can be made from heat foam plastic resin manufactured from metal mold of pushing out, the length of the solid-type pressure device is arbitrary. FIG. 4 is a six-corner honeycomb cell assembly. When two parts of the six-corner solid-type pressure device of half cut (12) are combined, they are shaped to be six-corner solid-type pressure device (14) with cylindrical vacant space (16) inside it. The six-corner solid-type pressure device (14) is long and slender object manufactured with arbitrary length. The air-type pressure device assembly (11) is stored in the cylindrical vacant space (16) of the six-corner solid-type pressure device (14). The six-corner honeycomb cell assembly (13) is manufactured by wrapping the six-corner solid-type pressure device (14) two or more times by the soft FRP prepreg (15). In this case, this soft FRP prepreg (15) becomes the external FRP wall of honeycomb cell of double wall. This six-corner honeycomb cell assembly (13) is used at the central position of honeycomb core assembly. At room temperature, the soft FRP prepreg (15) is the soft cloth, so it is not difficult to wrap six-corner solid-type pressure device (14) with the soft FRP prepreg (15). Because the adhesive of prepreg deteriorates at the room temperature, it is preferable to preserve the product within the freezer at minus 5° C. or less. FIG. 5 is a five-corner honeycomb cell assembly. When the six-corner solid-type pressure device of half cut (18) and the four-corner solid-type pressure device of half cut (19) are combined, they are shaped to be five-corner solid-type pressure device (21) with cylindrical vacant space (23) inside it. The five-corner solid-type pressure device (21) is a long and slender object with an arbitrary length. The air-type pressure device assembly (17) is stored in the cylindrical vacant space (23) of the five-corner solid-type pressure device (21). The five-corner honeycomb cell assembly (20) is manufactured by wrapping the five-corner solid-type pressure device (21) two or more times by the soft FRP prepreg (22). In this case, this soft FRP prepreg (22) becomes the external FRP wall of honeycomb cell of double wall. This five-corner honeycomb cell assembly (20) is used at the left and right side surface of honeycomb core assembly. At room temperature, the soft FRP prepreg (22) is the soft cloth, so it is not difficult to wrap five-corner solid-type pressure device (21) with the soft FRP prepreg (22). Again, because the adhesive of prepreg deteriorates at the room temperature, it is preferable to preserve the product within the freezer at minus 5° C. or less. FIG. 6 is a trapezoid filler assembly. The trapezoid filler assembly (25) is manufactured by wrapping the trapezoid-corner solid-type pressure device (24) with the soft FRP prepreg (26) two or more times. The trapezoid filler assembly (25) is a long and slender object and its length is the same as the honeycomb cell assembly. At room temperature, the soft FRP prepreg (26) is the soft cloth, so it is not difficult to wrap the trapezoid-corner solid-type pressure device (24) with the soft TRP prepreg (26). Generally, a honeycomb core is a set of many honeycomb cells of six-corner type and the external boundary of honeycomb core is trapezoid ruggedness surface. The trapezoid filler assembly (25) is used for correcting the ruggedness of top and bottom surfaces. In this case, this soft FRP prepreg (26) becomes the reinforcement structure of honeycomb structure and hull. Again, because the adhesive of prepreg deteriorates at the room temperature, it is preferable to preserve the product within the freezer at minus 5° C. or less. FIGS. 7A-7C show an assembly procedure explanation chart of FRP honeycomb core that illustrates the procedure sequence as follows: A. The process preparation is as follows: (a) The five-corner honeycomb cell assembly (28) and the trapezoid filler assembly (29) and the six-corner honeycomb cell assembly (27) preserved in the freezer at minus 5° C. or less are taken out from the freezer. (b) The adhesive function of FRP prepreg is lost at that temperature therefore it is not difficult to assemble them.B. The first step is as follows: (a) The five-corner honeycomb cell assembly (28) is placed on the left corner end. (b) The trapezoid filler assembly (29) is placed to next right. (c) The six-corner honeycomb cell assembly (27) is placed on the next position to the right. (d) As for clause (b) and clause (c), a necessary frequency is repeated according to the requested width of soft honeycomb core assembly. (e) At the end of the first step, the five-corner honeycomb cell assembly (28) is placed on the right corner end.C. The second step is as follows: (a) The five-corner honeycomb cell assembly (28) is placed on the left corner end of second step. (b) The six-corner honeycomb cell assembly (27) is placed on the next position to the right. (c) As for clause (b), a necessary frequency is repeated according to the requested width of soft honeycomb core assembly. (d) At the end of the second step, the five-corner honeycomb cell assembly (28) is placed on the right corner end.D. The third step is as follows: (a) Same work as the second step is repeated again till ahead one step to the last, according to the requested thickness of soft honeycomb core assembly (30).E. The last step is as follows: (a) The ruggedness surface of honeycomb core boundary is buried sequentially by using the trapezoid filler assembly (29). (b) After the boundary surface of soft honeycomb core assembly (30) is corrected from trapezoid ruggedness surface to the smooth surface, the soft honeycomb core assembly (30) is completed. Theoretically, the honeycomb structure of six-corner type can be infinitely arranged, and the structural position is unique. Therefore the honeycomb structure of an arbitrary size can be manufactured by repeating the similar procedure. Actually, it is not possible to arrange it infinitely because there is a size error margin in an individual solid pressure device assembly. When the epoxy system FRP prepreg is heated to the stiffening temperature, the air-type pressure device assembly and the solid-type pressure device assembly are expanded with heating. Therefore, it is desirable to design the size of the solid pressure device assembly smaller than the ideal shape. FIG. 8 is an assembly procedure explanation chart of FRP honeycomb structure. The soft FRP honeycomb structure (33) is manufactured from the soft FRP honeycomb core assembly (32) by bonding the soft FRP hull (31) on the top and bottom surface of the soft FRP honeycomb core assembly (32). The soft FRP hull (31) is made from two or more sheets of the soft FRP prepreg. The trapezoid filler assembly already corrects the top and bottom surfaces of the soft FRP honeycomb core assembly to the smooth surfaces, so the soft FRP hull is bonded on the wide and smooth surface of honeycomb core assembly. Therefore, the peel-off problem of honeycomb sandwich panel in the prior art is solved in this invention. FIG. 9 is an external frame assembly and soft FRP honeycomb structure. The process for stiffening process from the soft FRP honeycomb structure to rigid FRP honeycomb structure by heat and pressure is as follows: A. The external frame assembly (35) is manufactured by storing the soft FRP honeycomb structure (36) in the external frame (34). The external frame (34) is made from the frame structure parts and restrains all the surface of the soft FRP structure (36). B. The external frame assembly (35) is put in the heating oven. The external frame assembly (35) is heated during the fixed time in the heating oven at an appropriate temperature within the range of 100° C.-200° C. In general, the stiffening temperature of the epoxy adhesive is about 130° C. C. All the pressure device assemblies, the cylindrical air-type pressure device assembly (4) and the five-corner honeycomb cell assembly (20) and the six-corner honeycomb cell assembly (13) and the trapezoid filler assembly (25) are expanded by heating. By the way, all the soft FRP prepreg wrapping around solid-type pressure device is arranged to be honeycomb core assembly by the first process for manufacturing soft FRP honeycomb structure from soft FRP prepreg. The soft FRP prepreg is manufactured from adhesive resin and reinforced fiber. The adhesive resin in the soft FRP prepreg is melted by heating and is transformed plastically by internal pressure. Therefore, every soft FRP prepreg wrapping around all the honeycomb cells is bonded together permanently and is stiffened by heat and pressure. As the result, all the soft FRP prepreg materials become the honeycomb core assembly. A. The FRP honeycomb structure is manufactured by stiffening from the soft FRP honeycomb structure (36) to rigid FRP honeycomb structure (42) by heating at the same time as pressurizing by the reaction force between expansion pressure of the pressure device and the external frame (34). B. After enough cooling time, the rigid FRP honeycomb structure (42) is taken out of the external frame (34) to be the FRP honeycomb structure. FIG. 10 shows a rigid FRP honeycomb structure. The rigid FRP honeycomb structure (42) is composed of internal FRP wall (37), external FRP wall (38), six-corner foamed plastic resin (40), trapezoid-corner foamed plastic resin (41) and rigid FRP hull (39). Therefore, the rigid FRP honeycomb structure is characterized by double structural wall. A core assembly of the rigid FRP honeycomb structure is arranged parallel to the surface of honeycomb sandwich panel. The rigid FRP hull is bonded on the smooth and wide surface of honeycomb core assembly; therefore, the FRP honeycomb structure in this invention solves the peel-off problem. In the air tube, heat evaporation compounds such as alcohol and water remain in the honeycomb structure. It is desirable to abandon the heat evaporation compound by cutting the sealed ends. The process of manufacture of the honeycomb core assembly set tank characterized with double wall is almost the same as the process of manufacture of the FRP honeycomb structure in which honeycomb core assembly is arranged parallel to the surface. The process of manufacturing the honeycomb core assembly set tank is described herein below. FIG. 11 is the tank shape air pressure device (43) & tank shape air pressure device assembly (47). The tank shape air pressure device (43) is made from heatproof plastic material, it has arbitrary tank length (49) and it encloses the heat evaporation compound (44). The heat evaporation compound (44) is the heat blowing agent and the evaporating liquids. The evaporating liquids are water and alcohol. As for the internal atmospheric pressure, high pressure as possible is desirable if structural strength is permitted. The internal tank soft FRP wall (46) is for the reinforcement of the internal tank. The tank shape air pressure device assembly (47) is manufactured by the reinforcement procedure of the tank shape air pressure device (43) with soft FRP prepreg. The reinforcement procedure is as follows: 1. The soft FRP prepreg is cut into long and slender zonal. 2. Long and slender zonal prepreg is wrapping around the tank shape air pressure device (43). The size of the honeycomb core varies in proportion to the size of an internal tank. However, the manufacturing process after this is the same as the process of manufacture of the FRP honeycomb structure. The different point is that the vacant space shape, which is the internal shape of solid-type pressure device assembly, is changed from the cylindrical space to the tank shape space. The process of manufacture of the honeycomb core assembly set tank characterized with double wall is almost the same as the process of manufacture of the FRP honeycomb structure. The differences are the following 3 points: 1. cylindrical vacant space (16), (23)→tank shape space (45) 2. cylindrical air-type pressure device (3)→tank shape air pressure device (43) 3. air-type pressure device assembly (4)→tank shape air pressure device assembly (47) FIG. 12 is an example of honeycomb core assembly set tank. The heat foamed plastic resin (50) is made from solid-type pressure device by heating. The internal tank FRP wall (53) is made from the internal tank soft FRP wall. The external FRP wall (52) is made from the soft FRP prepreg wrapping around the solid-type pressure device. The rigid FRP hull of honeycomb core assembly set tank (51) is made from the soft FRP hull. The internal wall maintains the pressure of the internal tank. The external wall of the tank characterized by double wall endures the external shock loading. And the tank shape air pressure device assembly is manufactured from heatproof plastic material. The heatproof plastic tank is expanded by internal pressure and temperature. The shape of the internal tank is transformed permanently. When the set tank is taken out of the external frame, the expanding heat evaporation compound such as alcohol and water is remaining in the internal tanks of the set tank. It is desirable to abandon this heat evaporation compound by opening the shut off valve (54) of each internal tank. The specifications of the example of the honeycomb core assembly set tank are as follows: 1. The total capacity of set tank: 90 liters 2. Internal tank: (1) Diameter: (φ200 mm, (2) Length: 900 mm, (3) The number of internal tanks: 5 3. The size of set tank: (1) Length: 970 mm, (2) Width: 716 mm, (3) Height: 525 mm FIG. 13 is a honeycomb core assembly set tank stored in ISO 20-Foot container. Theoretically it becomes possible to manufacture a tank with infinite capacity. Because the honeycomb core structure composed by six-corner tanks has an infinite size. The capacity efficiency of the set tank is improved by increasing the number of accumulation of tanks of six-corner. Theoretically, there is no restriction in the length of the individual tank. The outline and the size of the honeycomb core assembly set tank stored in ISO 20-foot container are shown in FIG. 13. The specifications of the set tank are as follows: (1) ISO 20-foot container contains 2 units, (2) The capacity of one unit: 3000 liters and the number of internal tanks: 50, (3) The internal tank diameter: φ200 mm and the internal tank length: 1900 mm, (4) The total capacity ISO 20-foot container: 6000 liters, (5) The total number of internal tanks: 100 (50*2) When an internal tank is stiffened by reinforced fiber FRP, a newly developed high-pressure tank is manufactured by the heating and pressurizing device, which uses a mechanical reaction force of internal pressure devices with heat foam resin and external frame. Among the features of this system, the internal pressure device method uses the heating oven at ground atmospheric pressure. The heating oven is greatly cheap compared with autoclave because the strength is not needed for the partition wall of heating oven at ground atmospheric pressure. Internal capacity of the heating furnace is also larger than that of autoclave. The internal pressure method is the best for mass production. In addition, the internal pressure device method can manufacture a double wall tank where the external wall made of FRP protects the internal wall made of FRP. It is difficult to manufacture a double wall tank by autoclave. Further, the internal wall of a double wall tank maintains the high-pressure hydrogen gas, and the external wall protects a high-pressure tank from an external shock loading. Safety increases. In the internal pressure method, the stiffening pressure is strongly pressurized on to reinforced fiber FRP by both the vapor pressure power inside the tank and the expansion pressure of the heat foam resin. The structural test piece made thus completely cleared 30 tons, which is the maximum value of examination machine at the Yamanashi Prefecture Industrial Technology Center. Its strength corresponds to concrete material. In an embodiment of the invention, glass fiber may be substituted for carbon fiber. The price of the glass fiber corresponds to about 1/10 of prices of carbon fiber. It is likely to contribute to the reduction in costs of the high-pressure hydrogen tank used for the fuel cell powered vehicle. An embodiment of an on-board tank for the fuel cell powered vehicle is shown in FIGS. 20A-20D. The basic specs for such an on-board tank are as follows: 1. Hydrogen tank resisting pressure: 750 atmospheric pressures 2. Internal tanks (a) Capacity of single tank: 18.5 liters (b) Size of single tank: Internal tank diameter φ170 mm, the total length: 870 mm (c) Reinforcement thickness of internal tank: 10.0 mm (d) External wall reinforcement thickness: 11.4 mm3. A Set tank (a) Number of tank: 5 (b) Total capacity of set tank: 92 liters (c) Size of set tank: 528×716×970 mm (d) The total length of set tank (forecast): 1134 mm The high-pressure set tank manufactured according to the invention makes it possible that a lot of cylindrical internal tanks are arranged to be the six-corner honeycomb core assembly. The advantages are as follows: 1. A set tank can be built by one manufacturing process, which arranges a lot of single tanks to the honeycomb core assembly. (a) It saves the manufacturing cost. (b) The spacing efficiency of the tank arrangement can be improved.2. The hydrogen tank of the high pressure can be manufactured from the set of the double wall tanks. Therefore, structural strength to an external shock loading is reinforced.3. In general, a high-pressure tank is destroyed in the longitudinal direction. The bottom in the tank often drops off by the shock loading from the outside. Therefore, a high-pressure tank of the vehicle is put on transverse for the body. The set tank in the concept chart has been designed in the size that can be stored in the reception desk bonnet of the car. Because the set tank assembled with many small diameter tanks does not collapse in the direction of thickness, it is safe for driver.4. The external shape of the tank set becomes a rectangular hexahedron. Therefore, the method of fixing the hydrogen tank to the floor is easy. The hydrogen tank unit for transportation is shown in FIGS. 21A-21D. The basic specs for the tank unit are as follows: 1. Internal tank resisting pressure: 750 atmospheric pressures 2. Tank unit set (1 set=1 unit*2) (a) Unit capacity: 3000 liters (Set total capacity: 6000 liters) (b) Size of unit: 2652×2238×2500 (mm) (c) Size of ISO-20F container: 6096×2438×2591 (mm) (d) Number of internal tank: 50 (e) The process of manufacture: As well as the vehicle hydrogen tank The hydrogen tank unit for transportation is designed to be stored in the size of the ISO-20F container. The capacity of the tank unit shown in the concept chart is about 3000 liters. One unit is composed of 50 internal tanks and one set in ISO-20 container is composed of 2 units. The total capacity is 6000 liters. The advantages are as follows: 1. It can be expected that transportation at the room temperature becomes possible because the internal tank unit endures 750 atmospheric pressures. (a) The transport efficiency is the same as liquid gasoline because hydrogen gas is liquefied by the high pressure. (b) Even if the cooling power supply is lost by any chance, the safety is secured. It saves the cost for safety and maintenance.2. A tank unit can be built by one manufacturing process, which arranges a lot of single tanks to the honeycomb core assembly. (a) The spacing efficiency is improved by increasing the number of the internal tanks. (b) It saves the manufacturing cost.3. The hydrogen tank unit is manufactured from a lot of double wall tanks. Therefore, structural strength to an external shock loading is reinforced.4. The shape of the tank unit becomes a rectangular hexahedron. Therefore, the method of fixing the hydrogen tank unit to the floor is easy.5. Safety increases. Because a tank unit of many small tanks can divide the explosion energy to the sum of small energy. And, it is not easy to think about the situation to which all the small diameter tanks explode at the same time, when any trouble occurs. A transportation lorry for hydrogen tank unit is illustrated in FIGS. 2A-22B. The basic specs are as follows: 1. Lorry transportation (a) Lorry transportation: 4 unit (b) Lorry capacity: 12,000 liters (c) Lorry size: 14,992×2,838×3,481 (mm)2. Transportation temperature: Room temperature3. Internal tank resisting pressure: 750 atmospheric pressures The lorry transports two ISO-20F containers (4 hydrogen tank units). The advantages of this configuration are as follows: 1. The transportation efficiency is the same as liquid gasoline because hydrogen gas is liquefied in the high pressure. 2. The transportation cost is saved. The refrigerator for the liquid hydrogen is unnecessary on the lorry. Even if the cooling power supply is lost by any chance, the safety is secured. 3. Structural strength to an external shock loading is reinforced. The hydrogen tank unit is manufactured from a lot of double wall tanks and is protected by the container. An example of a hydrogen station is shown in FIGS. 23-24. The specification of the hydrogen station are as follows: 1. Size of site: 30×20(m) 2. Hydrogen tank unit: 8 unit (24,000 liters) 3. Special mention: The frozen power supply is unnecessary. In a preferred embodiment, the hydrogen station is designed to as to be operated at room temperature. The advantages of such an embodiment are as follows: 1. The tank unit of the hydrogen station can secure the structural strength with which the tank unit maintains 750 atmospheric pressures at room temperature. (a) The system that cools the hydrogen gas to the liquid hydrogen is not necessary at the hydrogen station. (b) When the cooling function is lost by any chance, the hydrogen gas need not be discharged in air. The safety is secured.2. The capacity of the hydrogen tank unit is 3000 liters per unit. (a) When several units of 3000 liters are prepared at the hydrogen station, it has the supply capacity equal to the gas station. (b) A large-scale underground tank is unnecessary. It will be possible to construct the hydrogen station cheaply than a present gas station. Moreover, the withdrawal is also easy. (c) It is expected that the maintenance cost is the same level as the gas station present.3. The liquid gasoline delivered by the tank lorry is transported to an underground tank at the gas station. It takes much time. The liquid hydrogen is kept in the storage warehouse on the ground because it is carried with the transportation unit. Much time is not necessary for unloading the transportation unit.4. It is possible to use the hydrogen unit tank for not only the hydrogen station but also the urgent power supply in public utilities such as the hospitals and subways. The destiny of the fuel cell powered vehicle is depending on whether it is possible to manage to make equal the equipment cost and the maintenance cost to the same of the gas station. Moreover, the safety is important. It will be appreciated that modifications may be made in the present invention. For example, the pressure device assembly, which is manufactured from heat foam plastic resin in this invention, can be manufactured from PET bottle or mold plastic. It will not be so difficult to make six-corner PET bottle, five-corner PET bottle and trapezoid-corner PET bottle. At that case, the honeycomb structure manufactured by that way will be a single wall cell structure with lightweight. The honeycomb cell structure of a single wall reinforced by FRP material is lightweight and it will be able to endure considerable weight. The spirit of this invention is horizontal arrangement of honeycomb core assembly. A honeycomb core assembly is the assembled structure from a lot of individual cells. For that purpose, this invention developed the manufacturing process of honeycomb core assembly of lightweight and high-strength. Accordingly, it should be understood that we intend to cover by the appended claims all modifications falling within the true spirit and scope of our invention. |
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043702988 | description | DESCRIPTION OF PREFERRED EMBODIMENT The following detailed description of the invention is intended to convey a clear understanding of the inventive concept but will be understood to be exemplary of other embodiments not specifically illustrated or described but within the scope of the invention as particularly set forth in the accompanying claims. Referring now to FIG. 1, there is illustrated a reactor system comprising a pressure vessel 10 in the shape of an ellipsoid having a substantially elliptically-shaped, in cross-section, wall 12 which forms an explosion chamber 14 for the containment of a series of fission explosions taking place at predetermined time intervals near the center O of the chamber. A vessel having the shape of an ellipsoid is preferred as it is best suited for containment of a nuclear explosion fireball which expands substantially spherically outwardly and which has a tendency to rise upwardly towards the top of the chamber. The pressure vessel must, of course, be absolutely safe and capable of withstanding the effects of a virtually unlimited number of fission explosions. More specifically, the vessel must be constructed of metals or metal alloys capable of withstanding severe conditions of blast, thermal shock, and chemical corrosion. The feasibility of constructing such a vessel for the containment of a series of explosions each having an energy release of about 100 tons of TNT, has been discussed as an eminently feasible project by R. J. Burke ANR Report (ENG-CTR TM-31). In the reactor system of the present invention, the shock and blast effects from each explosion are attenuated by injection of a quantity of a working fluid into the chamber immediately prior to each explosion. A liquid metal, preferably molten sodium, is used since, among sodium's many advantages, are its excellent heat transfer properties and the fact that its boiling point is high enough to allow the pressure vessel to operate at relatively low vapor pressures. In addition, sodium is a poor neutron moderator and has a low neutron capture cross-section. It is also contemplated to use lithium as the working fluid in the reactor system of the present invention since lithium has about three times the heat capacity per unit mass as does sodium. Accordingly, only about one-third as much power would be needed for pumping and recirculating lithium through the system as compared to molten sodium. Moreover, lithium has a much lower vapor pressure as compared to sodium so that there would be less alkali metal vapor produced within the chamber after each explosion. Working fluids, other than alkali metals, such as molten metals or alloys of lead, bismuth, magnesium, or tin can also be used. It would also be possible, with minor modifications, to use compounds such as sodium hydroxide as a working fluid in the reactor of this invention. The working fluid not only serves the purpose of protecting the vessel from the blast of the explosion but it also provides a carrier medium for the thermal energy generated by the explosion. Thus, the vessel must be constructed to withstand the blast effects of the working fluid as it is pushed outwardly with great impulse from the center of the vessel by the explosion to impact against the vessel wall. The vessel must also be designed to withstand a succession of thermal shocks which are produced as the wall is alternately heated and cooled by the working fluid. Specifically, as the working fluid is heated by an explosion, some of the thermal energy is transferred to the wall to thereby heat the wall. After the heated working fluid is withdrawn from the vessel, the wall is cooled again by the introduction of a quantity of cooler working fluid into the chamber immediately prior to the next explosion. The wall of the vessel must also be constructed of a material capable of resisting the chemically corrosive effects of the working fluid at a temperature of about 500.degree. C. Stainless clad, carbon steel or steel alloys such as are presently used in the manufacture of pressure vessels for conventional nuclear reactors are suitable for the construction of the pressure vessel of this invention. Other steel and steel alloys capable of resisting the thermal shocks in the reactor system of the present invention may be those presently used in the manufacture of gun barrels, or in the cylinders of diesel engines. Other metals such as Niobium, Molybdenum, Titanium, Tantalum, Tungsten, or alloys of such metals can be used as a cladding material since they have good physical properties and are highly resistant to the chemically corrosive effects of the working fluid. The selection of materials best suited for the construction of the pressure vessel, conduits, and the like, does not form a part of this invention, however. A supply system is provided which is particularly adapted for the introduction of the working fluid into the pressure vessel to obtain maximum attenuation of a fission explosion with a minimum detrimental effect on the integrity of the wall of the pressure vessel. It is estimated that it will be necessary to inject approximately 140 metric tons, if sodium is used, of the working fluid into the vessel for attenuating each explosion that has an energy release equivalent to about 4.times.10.sup.10 joules. Accordingly, the supply system includes a reservoir 16 which is constructed of a size capable of containing about 200 metric tons of working fluid. A plurality of auxiliary reservoirs C,D,E,F,G,H, and I are provided to supply the working fluid to select areas of the chamber. Thus, the auxiliary reservoir C is connected to the main reservoir 16 by means of a conduit 18. A high pressure pump 32 is provided in the conduit for continuously supplying working fluid to the auxiliary reservoir C. A branch conduit C.sub.31 is connected to the bottom of the auxiliary reservoir and to a passageway C.sub.11 extending through the wall of the vessel. A valve C.sub.1 is provided in the branch conduit C.sub.31 to interrupt the flow of working fluid to the passageway C.sub.11, on demand. Although only one branch conducts C.sub.31, valve C.sub.1 and passageway C.sub.11 is shown in FIG. 1, it will be understood that a greater number of the passageways C.sub.11 extend at spaced intervals through the wall 12 and circumferentially around the vessel so as to form a circular row of passageways. As more clearly shown in FIG. 2, a plurality of rows of passageways C.sub.11, C.sub.12 and C.sub.13 are preferably provided. The rows of passageways also extend at spaced intervals from the top of the vessel so that the uppermost portion of the vessel is provided with a multiplicity of spaced passageways for injection of the working fluid into the chamber. The auxiliary reservoir C is positioned at an elevation of about 20 meters above the vessel for delivering the working fluid at a relatively high pressure to the uppermost portion of the vessel and, more specifically, for delivering the working fluid immediately in front of the barrel opening 54. As in the case of the first auxiliary reservoir C, a second auxiliary reservoir C is continuously supplied with working fluid from the reservoir 16 through a main supply conduit 20 and high pressure pump 34. A branch conduit D.sub.31 is connected to the bottom of the second auxiliary reservoir D for supplying working fluid to a passageway D.sub.11 extending through the wall of the vessel. A valve D.sub.1 is provided in the branch conduit D.sub.31 to interrupt the flow of working fluid to the passageway, on demand. As more clearly illustrated in FIG. 2, a plurality of rows of circumferentially extending passageways D.sub.11 -D.sub.14, are provided. Each branch conduit is provided with a valve D.sub.1 -D.sub.4 to interrupt the flow of working fluid to each passageway, on demand. The second auxiliary reservoir is positioned at an elevation of about 6 meters above the passageways to provide working fluid to the passageways at a pressure which is somewhat lower than the pressure in passageways C.sub.11 -C.sub.13. A third auxiliary reservoir E is connected by means of a main supply conduit 22 to the main reservoir 16. A high pressure pump 36 is connected in the main supply conduit to continuously supply working fluid to the third auxiliary reservoir. As shown in FIGS. 1 and 2, a plurality of circumferentially extending rows of passageways E.sub.11 -E.sub.15 extend through the wall of the vessel and each passageway is supplied with working fluid through a branch conduit E.sub.31 connected to the bottom of the third auxiliary reservoir E. A valve E.sub.1 -E.sub.5 is provided in each branch conduit to supply working fluid to each of the passageways, on demand. The third auxiliary reservoir is positioned at an elevation of about 3 meters above the passageways E.sub.11 -E.sub.15. Additional auxiliary reservoirs F, G and H are provided to supply working fluid through passageways F.sub.11, G.sub.11 and H.sub.11, respectively, and in the form of a spray, shown at F.sub.21, G.sub.21 and H.sub.21, to the chamber. Each of the auxiliary reservoirs F, G and H is connected to the main reservoir 16 by means of main supply conduits 24, 26 and 28. A high pressure pump 37, 38 and 40 is connected in each of the main supply conduits for continuously supplying working fluid to each of the auxiliary reservoirs F, G and H. Each of the auxiliary reservoirs F, G and H is also provided with branch conduits F.sub.31, G.sub.31 and H.sub.31 connected to each of the passageways F.sub.11, G.sub.11, and H.sub.11, respectively. A valve F.sub.1, G.sub.1 and H.sub.1 is connected in each branch conduit to interrupt the flow of working fluid to each of the passageways, on demand. The auxiliary reservoirs F, G and H provide working fluid at relatively low pressure to the rows of passageways F.sub.11, G.sub.11 and H.sub.11 and, accordingly, these reservoirs are positioned about 2 meters above their respective rows of passageways. The passageways F.sub.11, G.sub.11 and H.sub.11 extend over a major arcuate portion of the chamber wall and inject the working fluid part way into the chamber prior to an explosion in the form of a fan or cone shaped spray, shown at F.sub.21, G.sub.21 and H.sub.21. The low pressure spray from the passageways F.sub.11, G.sub.11 and H.sub.11 is primarily intended for protecting the wall of the chamber from the effects of the explosion, but it also has the additional beneficial effect of cooling the wall of the vessel to maintain the temperature of the wall at a substantially constant temperature to thereby minimize thermal shocks to the wall. The auxiliary reservoir I is provided to supply working fluid, under relatively high pressure, to the bottom portion of the vessel and especially in front of the barrel opening 54. A main supply conduit 29 is connected to the main reservoir 16 and a high pressure pump 42 is connected in the main supply conduit 29 for continuously supplying the working fluid from the main reservoir to the auxiliary reservoir I. A branch conduit 30 is connected between the auxiliary reservoir I and each of the passageways extending through the wall of the vessel for supplying working fluid to the passageways. A valve is connected in each branch conduit 30 to interrupt the flow of working fluid to the passageway on demand. As more clearly illustrated in FIG. 3, three circumferential rows of passageways I.sub.11, I.sub.12 and I.sub.13 are provided to supply the working fluid as a spray I.sub.21, I.sub.22 and I.sub.23 in a select area at the bottom of the vessel. Each branch conduit is controlled by a valve I.sub.1, I.sub.2 and I.sub.3 to interrupt the flow of working fluid, on demand. The passageways associated with the auxiliary reservoirs C, D, and I each have a diameter of about 1 cm. and occupy about 40% of the volume of the wall through which they extend. Thus, when viewed from the inside of the vessel, the openings provided by the passageways C.sub.11 -C.sub.13 and D.sub.11 -D.sub.14 will occupy about 40% of the area over which they extend. The passageways associated with the auxiliary reservoir E each have a diameter of about 2 cm and occupy about 60% of the volume of the wall portion through which they extend. The passageways associated with the auxiliary reservoirs F, G, and H each have a diameter of about 1 cm. and occupy about 20% of the volume of the wall portion through which they extend. Relatively small amounts of the working fluid are supplied to the passageways from each of the auxiliary reservoirs C, D, F, G, H and I so that these reservoirs can be constructed of a relatively small size. Since auxiliary reservoir E, however, will supply as much as 70% of the working fluid to the chamber, this reservoir will be constructed of relatively larger size capable of holding about 100 metric tons of the working fluid. The auxiliary reservoirs can be constructed in the form of toroidal chambers centrally positioned with respect to a central longitudinal axis extending vertically through the center of the vessel. Alternatively, the auxiliary reservoirs can be constructed as cylindrically shaped vessels or containers positioned at circumferentially spaced intervals of 60.degree. to 120.degree. around the vessel. However, for maximum utilization of space, it is preferred to construct the auxiliary reservoirs as toroidal chambers mounted at spaced intervals from the top to the bottom of the vessel. A pair of slugs A and B are illustrated in FIG. 1 as they move toward each other for interception near the center 0 of the vessel. It is preferred to have the slugs intercept at a point slightly below the center in view of the tendency of the fireball from the explosion to rise towards the top of the chamber. Since the vessel is designed to contain fission explosions equivalent to about 9.5 tons of TNT, the chamber 14 has a major internal radius of about 12 meters along a vertical axis and a minor internal radius of about 10 meters along a horizontal axis. The greatest amount of attenuation to the explosion will be obtained when the working fluid is as closely concentrated near the point of interception of the two slugs, as possible. Thus, a slight variation in the point of interception of the slugs above or below the intended point of interception is not critical and would merely result in a slight decrease in the degree of attenuation of the explosion. The nearer the center of concurrence, which is the center of fissioning, of a pair of slugs is to the geometric center of the concentrated working fluid, the greater will be the magnitude of the resulting explosion. This is because the atoms of the concentrated working fluid act somewhat as neutron reflectors. Since the working fluid is a slurry which contains fissile actinides and other actinides which may fission to faster neutrons some neutrons will be generated within this spray. Likewise, if the concurrence of slugs is "off center" the resulting explosion will be of smaller magnitude. A pair of accelerating mechanism 50, only one of which is shown in FIGS. 1 and 3, are provided at diametrically opposite ends of the vessel for propelling the slugs A and B into the vessel. Each accelerating mechanism is designed to propel a slug through a barrel or passageway 52 and through a barrel opening 54 provided in the vessel 10 into the chamber 14. A closure mechanism, which will be described in greater detail in connection with FIG. 3, is provided between the barrel 52 and the barrel opening 54 for sealing the barrel opening and the chamber at the instant of an explosion. Accordingly, the closure mechanism must be constructed to prevent leakage of radioactive solid and gaseous materials from the chamber at all times, to prevent contamination of the surrounding area. It is particularly desirable to protect each barrel opening 54 and closure mechanism from the effects of the explosions and, accordingly, the working fluid is preferably concentrated as a spray immediately in front of each barrel opening. Since the auxiliary reservoir C is located at an elevation of about 20 meters above the passageways C.sub.11 -C.sub.13, the static pressure acting on the working fluid will force the working fluid through the passageways at great pressure. It is also desirable to prevent the working fluid from impinging upon a slug as it passes out of the barrel opening 54 into the chamber and as it moves toward the center of the chamber. Since the slug travels at a speed of about 50 to 300 m/sec. only a very short time interval is available for injecting the working fluid from the first auxiliary reservoir C in front of the barrel opening from the instant that a slug leaves the barrel opening. Thus, the working fluid may additionally be placed under positive pressure by providing a pressure pump in each branch conduit D.sub.31 upstream of the valve D.sub.1 to force the working fluid in front of the barrel opening at the instant that the slug passes out of the barrel opening into the chamber 14. To provide additional protection to the closure mechanism, the vessel is also provided with an elongated neck 58 at the upper and lower ends of the vessel which has the effect of lengthening the barrel openings 54 to about 3 to 4 meters to thereby increase the distance that a shock wave from an explosion must travel. Thus, the blast from the explosion is substantially reduced as it travels outwardly from the center of the chamber toward the wall of the vessel and into the barrel openings 54. An innermost row of passageways E.sub.11 connected to the third auxiliary reservoir E extend vertically through the wall and surround the major vertical axis of the vessel to provide an inner curtain E.sub.21 of working fluid which falls through the chamber and which has a somewhat tubular shape extending from the top to the bottom of the chamber. The inner curtain therefor forms a somewhat tubular passage for the slugs A and B having an inner radius from the center of the chamber of about 40 cm. Although only the innermost row of passageways E.sub.11 are shown to provide the tubular curtain E.sub.21, it is to be understood that a plurality of concentric rows of such vertically oriented passageways may be used to provide the tubular curtain of working fluid of somewhat greater thickness and density. In other words, a plurality of concentric rows of passageways extending over a radial distance of from 40 to 50 cm. from the central longitudinal axis may be utilized in the formation of the tubular curtain. The valves E.sub.1 which control the flow of working fluid forming the inner tubular curtain are activated over a longer time interval as compared to the remaining valves E.sub.2 -E.sub.5 so that the inner tubular curtain is allowed to extend over the entire length of the chamber along the major vertical axis of the vessel before an explosion takes place. Successive rows of the passageways E.sub.12 -E.sub.15, outwardly of the innermost row of passageways E.sub.11, are under the control of the valves E.sub.2 -E.sub.5 to provide a plurality of concentric tubularly shaped curtains E.sub.22 -E.sub.25 of the falling working fluid which are of a shorter length than the innermost curtain E.sub.21 or which are preferably of a gradually decreasing length the greater the radial distance of the rows of passageways E.sub.12 -E.sub.15 from the major longitudinal axis of the vessel. The effect of the concentric curtains is such that at the instant of the explosion preferably about 70 to 80% of the working fluid is concentrated near the center of the chamber and in a somewhat spherical pattern, somewhat as illustrated in FIG. 1. Stated in another way, the concentric tubular curtains of falling working fluid are of a progressively shorter length from an innermost curtain to the outermost curtain to thus form a somewhat spherical pattern of the working fluid closely adjacent to the center of the vessel and to the point of interception of the slugs. The radial distance of the outermost row of passageways E.sub.15 is located about 4.5 to 5 meters from the major longitudinal axis of the vessel. However, to provide for a greater concentration of the working fluid around an exploding pair of slugs, the passageways E.sub.12 -E.sub.15, with the exception of the vertical passageways E.sub.11, are inclined slightly inwardly toward the center 0 of the chamber 14 so that with the outer row of passageways E.sub.15 having an inclination which is slightly greater than the inclination of the inner row of passageways E.sub.12. Accordingly, the falling curtains E.sub.22 -E.sub.25 have a tendency to fall inwardly toward the inner curtain E.sub.21, to thereby converge upon the inner curtain somewhat as illustrated in FIGS. 1 and 2. The converging curtains form a denser mass of the working fluid at the center of the chamber to thereby increase the degree of attenuation to the explosion and to form the spherical pattern produced at the center of the chamber around the inner curtain E.sub.21 with the spherical pattern having an outer radius of about 31/4 meters from the center of the chamber to the outer tubular curtain E.sub.15. In order to produce the spherical pattern of the working fluid at the center of the chamber, it is necessary to activate the valves E.sub.1 -E.sub.5 in a sequence whereby the innermost valves E.sub.1 are activated first and for the longest time interval to permit the working fluid to flow through the passageways E.sub.11 for a duration long enough to form the vertical tubular curtain extending along the entire length of the vessel immediately prior to the explosion. Thus, the innermost valves E.sub.1 are activated first followed in sequence by the valves E.sub.2 to E.sub.5 in each concentric row outwardly of the innermost row of valves E.sub.1. Each succeeding row of valves from the outermost to the innermost rows of valves are thereafter closed to complete the formation of the tubular curtains E.sub.22 to E.sub.25 of gradually decreasing length. The valves E.sub.5 in the outermost row are therefor open for the shortest duration of time to thereby form the relatively short tubular curtain E.sub.25 of working fluid. For purposes of illustration, only five concentric rows of the passageways E.sub.11 -E.sub.15 are shown in FIGS. 1 and 2. It will be understood, however, that a greater number of rows of passageways may be provided to assure that between 70-80% of the working fluid flows through the passageways to be concentrated at the center of the vessel at the instant of an explosion. In terms of volume, this means that about 70 to 80 tons of working fluid, if sodium is used, will be concentrated in less than 3% of the total volume of the chamber. These values can be suitably adjusted if, for example, it is desired to supply a greater amount of the working fluid near the chamber walls from the reservoirs F, G and H. This can be accomplished by actuating the valves F.sub.1, G.sub.1 and H.sub.1 for a longer duration of time while the valves E.sub.1 -E.sub.5 are actuated for a shorter duration of time to thereby provide a smaller amount of the working fluid of perhaps 50-60% or 50-60 tons to the center of the chamber. Although the auxiliary reservoirs F, G and H have a static head of only about 2 meters, the valves F.sub.1, G.sub.1 and H.sub.1 can be activated for a longer time period if it is desired to provide an increased attenuation to the explosion as well as an increase in the cooling of the wall 12. The passageways F.sub.11, G.sub.11 and H.sub.11 are generally directed towards a point slightly above the center of the vessel. In view of the relatively large number of passageways which are concentrated at the top and at the bottom of the vessel, it is necessary to provide additional strength to the top and bottom portions of the vessel. Accordingly, the wall 12 is formed of a gradually increasing thickness from a central horizontal portion of the vessel where the wall is about 1 meter thick, towards the upper and lower ends of the vessel where the wall is about twice as thick, or about 2 meters. The total number of passageways in the vessel is dependent upon the diameter of each passageway and on the pressure of the working fluid flowing through the passageways. Accordingly, the larger the diameter of each passageway and the greater the pressure of the working fluid, the greater the amount of working fluid which will be injected into the chamber over a given period of time. The lower portion of the vessel, illustrated in FIG. 3, forms a circular collection through or sump 60 for the collection and drainage of the heated working fluid from the chamber 14 after each explosion. A projection 62 extending inwardly into the chamber along the major vertical axis of the vessel is provided to prevent the flow of working fluid into the barrel opening 54. The barrel opening 54 extends centrally through the projection 62 so that the mouth of the barrel opening, formed as a rounded surface 63 on the projection, is positioned at a distance of about 1 to 2 meters above a bottom surface 64 of the trough, as shown. Except for a relatively small amount of the working fluid which is sprayed immediately in front of the barrel opening 54, the greater portion of the working fluid from the tubular shaped curtains E.sub.21 -E.sub.25 will surround the projection 62 without the working fluid actually falling into the barrel opening 54. The projection 62 must be protected from the effects of the explosions and, for this purpose, the passageways I.sub.13 -I.sub.15 extend into the projection where they terminate in the rounded surface 63, as shown in FIG. 3. The working fluid flows through the passageways I.sub.13 only during the short time interval that a slug travels from the mouth of the barrel opening 54 to the point of interception near the center O of the chamber. Since the outwardly and upwardly directed spray, shown at I.sub.21 and I.sub.22 do not intercept the flight path of a slug, the valves I.sub.1 and I.sub.2 can be opened for a longer period of time and before a slug leaves the barrel opening 54. Accordingly, the passageways I.sub.11 and I.sub.12 could be connected to the auxiliary reservoir H which would provide working fluid to these passageways at a static pressure of only about 7 meters which would be sufficient to supply an adequate quantity of the working fluid to the projection 62 at the instant of an explosion. A plurality of outlet passageways 70 extend from the bottom surface 64 of the trough 60 through the wall 12 for continuously draining the working fluid from the chamber. Although only one passageway is shown in FIGS. 1 and 3, it will be understood that any number of such passageways can be provided as long as the passageways are of a size capable of draining all of the working fluid from the vessel between explosions. Five to nine working fluid outlet passageways are provided to obtain complete drainage of about 150 tons of the working fluid from the vessel between explosions. Of course, if only 3 passageways are provided, the passageways must be of a relatively large size. The spaces between each pair of outlet passageways 70 at the bottom surface 64 of the trough are provided with an inwardly sloping surface or with a crest, not shown, so that the working fluid will drain out of the trough into the passageways without leaving any pockets in which the working fluid can collect and which would cause a gradual build-up of radioactive debris in the pockets. A major portion of the debris which is produced by the exploded slugs, and which is entrained within the working fluid, is passed out of the vessel and is partially separated from the working fluid in a catch basin 72 which is provided for each outlet passageway 70. Alternatively, a toroidally shaped catch basin is preferably provided extending horizontally around the lower portion of the vessel with all of the outlet passageways 70 connecting into the toroidal basin. The working fluid, heated after each explosion, may be used to pump a liquid in a manner described in Application Ser. No. 383,828, filed on July 30, 1973 by Edward F. Marwick. Alternatively, the heated working fluid may be utilized in the generation of electric power. Neither method of performing useful work form a part of the present invention and are therefore not described in this application except insofar as to show how the heated working fluid is circulated through the reactor system. The working fluid is conducted out of the catch basin 72 through a conduit 74 and into a heat exchanger 76. A heat exchange fluid is passed, in indirect heat exchange relationship with the heated working fluid, through a heat exchange coil 78 positioned within the heat exchanger. Inlet and outlet conduits 80 are connected to the coil 78 and to a heat engine means which could drive an electric generator (not shown). A conduit 82 is connected between the heat exchanger 76 and the main reservoir 16 for conducting cooled working fluid to the main reservoir. The closure mechanism for closing the lower barrel opening 54 is more clearly illustrated in FIG. 3. Since the closure mechanism at the top of the vessel is identical in construction it will not be described again. The closure mechanism may take the form of a hinged, single or double panel, trap door, or a shutter mechanism based on the principle of a camera shutter. It is preferred, however, to provide a closure mechanism consisting of a sliding or rotating plate having one or more openings. In the case of a rotating plate, the plate may be rotatably mounted on an axis offset from the central longitudinal axis of the vessel. Stepwise rotation of the plate would successively bring an opening in the plate into registration with the barrel opening 54 and the barrel 52. A main closure plate 90, which for purposes of simplicity is shown as a sliding plate, is shown in a position where an opening 92 is in alignment with the barrel opening 54 in the vessel. Various means for propelling a slug into the chamber may be employed such as chemical explosives, for example, such as are commonly used for propelling projectiles through a gun barrel. However, most chemical explosives use oxygen compounded with nitrogen so that oxygen from the chemical explosion products, if introduced into the chamber, would readily combine chemically with an alkali metal working fluid, such as liquid sodium. The resulting oxide in the liquid sodium would make the working fluid extremely corrosive. Moreover, nitrogen from the chemical explosion products has a cross-section to slow neutrons of 1.88 barns and thus would absorb neutrons produced in the explosion and hence reduce the breeding ratio for breeding valuable isotopes in the reactor system of the present invention. If a slug accelerating medium such as hydrogen gas is used, problems would arise due to the fact that hydrogen forms hydrides with alkali metals and, at higher temperatures, the hydride formation is also extremely corrosive and could cause damage to some refractory metals used in the structural components of the system. Another type of slug accelerating mechanism could employ a piston for compressing a quantity of gas in a compression chamber. The compressed gas is then suddenly released into the barrel to rapidly propel a slug through the barrel into the chamber. Other types of accelerating mechanism could be of mechanical construction operating on the principle of a cross-bow or catapult. Any mechanically operated accelerating mechanism would have the advantage that it would not introduce extraneous matter into the chamber, such as solids or gases produced by chemical explosives. In the preferred method a slug is accelerated by gas pressure in which a more suitable gas, such as helium would be used. The helium gas would be subject to the application of a controlled quantity of heat energy obtained by the discharge of an electric charge of predetermined intensity into a pressure chamber containing the helium gas under pressure. Instead of discharging an electric charge into the pressure chamber, the helium gas can also be rapidly heated by subjecting the gas to microwaves. By controlling the intensity of an electrical discharge from a capacitor, for example, or by controlling the intensity or duration of the microwaves, it is possible to obtain a fine control over the amount of heat energy supplied to the gas to thereby control the amount of pressure of the heated gas in the pressure chamber and ultimately the speed of a slug. Since a plurality of slugs are simultaneously propelled into the chamber, an electrical control system, under the control of a computer can be provided to exercise accurate control over the acceleration of the slugs to assure that they will intercept at a predetermined speed near the center of the chamber. A slight variation in the speed of the slugs of about one meter per second would not be critical where the slugs are ultimately accelerated to velocities of about 50 to 300 m/sec. Calculations will show that a variation of about one meter per second in the speed of the slugs would mean that the slugs would intercept only a few centimeters above or below the intended point of interception in the chamber. Moreover, the vessel is constructed of sufficient strength and with a safety factor capable of withstanding explosions which are several times greater in intensity than the intensity of explosions contemplated in the present invention. The use of helium gas as a propellent is well within the present state of the art since, over the last few years, much work has been done in the development of high-velocity gas guns. Such guns could readily be adapted for use in propelling the slugs of the present invention. Reference is made to one development in this field by J. D. Watson in "A Summary of the Development of Large Explosion Guns for Re-Entry Simulation" (AD-720394; PIFR155) in which the author describes a program to develop an explosive gun capable of launching large saboted models to re-entry velocities. Saboted lithium-magnesium models of up to 4.5 inches in diameter were launched successfully to a speed of 4.8 km/sec. A velocity record was also established by accelerating a 2-gram cylindrical projectile to a velocity of 12.2 km/sec. Another high velocity gas gun development is described by the authors G. R. Fowles et al., in "The Review of Scientific Instruments" Vol. 41, No. 7 of July, 1970. The slug accelerating mechanism 50 is intended to be of the type for rapidly heating and expanding a quantity of helium gas within an expansion chamber. Alternatively, one extreme end portion of the barrel 52 itself could form an expansion chamber for the helium gas. The size of the expansion chamber behind the slug can be designed to take into account slugs of different weights or slugs which are accelerated to a particular desired velocity. Moreover, the amount of pressure of the gas in the expansion chamber and the intensity of the electrical discharge can also be varied to accelerate a slug to the desired velocity. A helium storage and supply tank 92 is provided to supply a predetermined quantity of the helium gas to the expansion chamber in the accelerating mechanism or, alternatively, directly to the expansion chamber in the end portion of the barrel 52. The helium gas is supplied to the accelerating mechanism at the desired pressure by means of a pump 94 positioned in conduit 96 connecting the storage tank 92 to the accelerating mechanism 50. A check valve 98 is provided in the conduit 96 to close the conduit as soon as a predetermined quantity of the helium gas has been supplied to the accelerating mechanism. The slugs which are formed of a sub-critical mass of compressed actinides are manufactured in a slug manufacturing station 100 and conveyed into the breech (not shown) of the barrel 52. The insertion of a slug into the breech of the barrel must take place rapidly if it is desired to produce an explosion every 5 to 10 seconds. Several mechanisms for accomplishing this are briefly described, but since such mechanisms are well known in the art and do not form any part of this invention, they are not specifically illustrated in the drawings. One mechanism for rapidly injecting a slug into the breech may be a pusher mechanism for injecting a slug through an opening in the breech. The opening can be closed and sealed by means of a hinged or sliding panel on the barrel. Alternatively, a revolver-type magazine can be provided which is rotatably mounted around the barrel and which has a plurality of chambers, usually six, for the slugs. An injection mechanism operating in conjunction with the rotatable magazine would rapidly inject a slug out of a magazine chamber and into the breech of the barrel. In another method, a plurality of barrels, in alignment with a corresponding plurality of barrel openings 54 may be provided for sequentially propelling slugs into the chamber. This method would have the advantage that if a barrel should need to be repaired or replaced, the remaining barrel or barrels would continue to operate and propel slugs into the chamber without any interruption in the operation of the reactor system. In another system, a plurality of barrel sections of relatively short length of perhaps 50-100 cm. would be fixedly mounted in a frame which is rotatable on a central longitudinal axis. Each barrel section would move into registry and alignment with the barrel 52 to sequentially position a slug in the barrel section in alignment with the barrel above the expansion chamber. Here again, failure of one of the barrel sections would not result in a shutdown of the entire reactor system since the remaining barrel sections would continue to function in positioning slugs in the remaining barrel sections into alignment with the barrel for propelling the slugs into the chamber. Each slug is cylindrical in shape and has a slightly larger external diameter than the internal diameter of the barrel 52 so that it will be held in position in the barrel by frictional force before it is propelled into the chamber. The barrel 52, at the bottom portion of the vessel is provided with a plurality of vent openings 102 extending over a portion of its length adjacent to the barrel opening 54. The vent openings extend through the wall of the barrel and lead into a main expansion chamber 104 and an auxiliary expansion chamber 104'. The expansion chambers extend between an exterior surface of the barrel and a surrounding main gas containment jacket 106 and auxiliary gas containment jacket 106', respectively. The combined volume of these expansion chambers should preferably be at least about 20 times the volume of the barrel which it surrounds so that a major portion of the hot expanding helium gas, of at least about 95%, used for propelling a slug through the barrel will be dissipated into the expansion chambers. A cryogenic pump, not shown, may be provided in the chambers to cool and condense the helium gas. The gas condensate is then readily withdrawn from the expansion chambers and recirculated to the helium storage and supply tank. Alternatively, and as shown in FIG. 3, the helium gas is evacuated from the expansion chambers 104 and 104' through an opening 110 in each jacket and returned to the helium storage and supply tank 92 through a return conduit 108. A vacuum pump 112 is connected in the conduit 108 for the continuous evacuation of the expansion chambers and barrel 52. A rotatable or slidable plate 114, of the same construction as the main plate 90, is preferably provided between the gas containment jackets 106 and 106'. Thus, the auxiliary expansion chamber 104 is of a relatively short length extending between the plates 90 and 114 while the main expansion chamber 104 is of a relatively long length extending partway between the auxiliary plate and the accelerating mechanism 50. The auxiliary plate 114 is shown in a position where an opening 116 in the plate is out of alignment with the barrel 52 to thereby isolate and seal the auxiliary expansion chamber 104' from the main chamber 104. In the operation of the slug accelerating mechanism 50 and before a slug is accelerated to the desired velocity by the expanding helium gas in the expansion chamber the main and auxiliary plates 90 and 114 are both in an open position so that the openings 92 and 116 in the plates are in registration with the barrel 52 and barrel opening 54. As a slug, here shown by reference character B, passes the auxiliary plate 114, an actuator 116 moves the auxiliary plate to the closed position, as illustrated in FIG. 3, to seal the auxiliary expansion chamber 104' from the main expansion chamber 104. The rapidly expanding helium gas in the barrel 52 is vented through the vent openings 102 into the main expansion chamber 104 to be evacuated by vacuum pump 112 and to be returned through conduit 108 to the helium storage and supply tank 92. The helium gas is repressurized, cooled and stored in the storage tank for re-use in propelling additional slugs into the chamber. As the slug B continues to travel through the barrel, it passes through the opening 92 in the main plate 90 and is propelled through the barrel opening 54 into the chamber. At the instant the slug B passes the main plate 90, the actuator 116 moves the main plate 90 to the closed position. The chamber 14 is therefore sealed by the main plate at the instant of an explosion in the chamber. The main plate 90 is moved by the actuator 116 to an open position immediately following an explosion in the chamber 14 while the auxiliary plate 114 remains closed for a short interval after an explosion. Solid and liquid materials such as the working fluid and radioactive debris from the exploded slugs as well as gaseous materials such as helium gas which enter the barrel between the plates 90 and 114 at the instant that the main plate 90 is opened, are expelled through the vent openings 102 into the auxiliary expansion chamber 104'. A branch conduit 118 extends from the opening 110 at the lower part of the auxiliary expansion chamber 104' to a separator 120 in which the solid and liquid materials are separated from the gaseous materials. The vacuum pump 112 is connected by conduit 108 to the separator 120 for continuously withdrawing the helium gas from the separator and for returning the helium gas to the storage tank 92. The working fluid and radioactive debris from the exploded slugs are recirculated to the main reservoir 16 or directly to the slug manufacturing station 100 through a conduit 122 leading from the separator 120. The length of the auxiliary jacket 106' between the plates 90 and 114 is about one-tenth the length of the main jacket 106 since the auxiliary expansion chamber 104' is primarily intended for the purpose of removing the solid and liquid materials which have fallen into the barrel opening 54 and for evacuating gaseous materials from the chamber 14. Helium gas escaping past the auxiliary plate 114 during the time that the slug B is propelled through the opening 116 in the auxiliary plate 114 and during the time that the main plate 90 is moved to the closed position is also evacuated from the auxiliary expansion chamber 104'. Immediately following the removal of solid, liquid and gaseous materials from the auxiliary expansion chamber, the auxiliary plate 114 is moved by the actuator 116 to the open position in preparation for propelling the next slug into the chamber. During that time, the vacuum pump 112 continues its operation to evacuate additional amounts of gaseous materials including helium, krypton and xenon, and alkali metal vapors from the chamber 14. Accordingly, substantially all of the helium gas used for propelling slugs in continuously evacuated from the barrel 52 and chamber 14, except for trace amounts which are entrained in the working fluid and which are drained out of the chamber into the sump 72, through the expansion chambers 104 and 104' so that there is never more than a residual amount of the helium gas in the chamber to conduct the undesired transmission of shock waves to the walls of the chamber. The closure mechanism at the top of the vessel is substantially of identical construction as the closure mechanism at the bottom of the vessel. However, since gravity causes the solid and liquid materials to fall to the bottom of the vessel, only trace amounts of these materials in the form of dust, spray or vapor will enter the barrel opening 54 or the barrel 52. Accordingly, the auxiliary plate 114 can be omitted and, instead, a single gas containment jacket 106 can be provided for the evacuation of these materials and gases out of the chamber 14 and the barrel 52. Any helium gas so withdrawn is separated from other radioactive gases and solid or liquid materials and returned to the helium storage tank 92. It is important to note that the accelerating mechanism 50; the closure mechanism including the actuator 116, the sliding plates 90 and 114, and the gas containment jackets 106 and 106' are all positioned externally of the vessel. Thus, all of these components can be used for a practically unlimited number of times and until such time as natural wear and tear requires replacement of any of these components. Accordingly, the only component of the reactor system which is subject to the effects of the explosions is the vessel itself and the main sliding plates 90 at the opposite ends of the vessel. These are, however, designed with a safety factor to assure safe performance for a virtually unlimited number of explosions. The following parameters in the operation of the present reactor system are selected for illustrative purposes and can be suitably varied without detracting in any way from the overall operation of the system. Accordingly, with the reactor system operating on a 6 second cycle, that is, with an explosion taking place every 6 seconds and with the slugs traveling at a speed of 100 m/sec. as they approach interception near the center O of the vessel, the operation of the reactor system is somewhat as follows. At the beginning of the cycle, or at 0 seconds, the slugs are accelerated at about 50 m/sec.sup.2. At 2.0 seconds, the auxiliary sliding plates 114 are open and the slugs have reached the desired velocity of about 100 m/sec. Of course, the slugs no longer accelerate once they pass the vent openings 102 in the barrels 52. At about 2.02 seconds, the slugs pass the auxiliary plates 114. At 2.04 seconds, the slugs pass the main plates 90 and the auxiliary plates 114 begin to close. At 2.06 seconds, the main plates 90 begin to close and the slugs enter the chamber at about 12 meters from the center O. The main and the auxiliary plates will be closed at the instant of the explosion which takes place at 2.18 seconds. At about 3.2 seconds, the main plates 90 are opened to evacuate solid, liquid and gaseous materials from the chamber 14 and through the barrel openings 102 and the auxiliary expansion chambers 104'. The fission products produced by the explosion include gases such as krypton (Kr) and xenon (Xe), volatile fission products such as Rubidium (Rb) and Cesium (Cs), and microscopic particles of actinides. These explosion products are separated from the working fluid and the helium gas in the separator 120 by methods well known in the art and will not be described in further detail. During the time interval that the main plates 90 are opened at 3.2 seconds and the end of the cycle at 6 seconds, the working fluid is allowed to drain out of the chamber through outlet openings 70 into the catch basin 72, new slugs are inserted into the breech of each barrel 52, and helium gas is charged into the expansion chambers in the accelerating mechanisms or into the ends of the barrels 52 in preparation for commencement of the next cycle. The outlet passageways 70 are always open which means that during the entire 6 second cycle there is a continuous drainage through the passageways into catch basin 72. In the above illustrative example, the acceleration of a slug takes place at a constant 50 m/sec.sup.2 so that the final velocity of the slug at a point of interception near the center of the chamber is 100 m/sec. The advantages of a relatively low rate of acceleration of 50 m/sec.sup.2 are that there are less disruptive forces acting upn the slug during the flight and that there is more time to control its acceleration. Moreover, less helium gas from the accelerating mechanism will enter into the chamber. Of course, the length of the barrel will therefore be about 100 meters. If a higher rate of acceleration is used the length of the barrel and the time of accelerating the slug will be less. For example, if the rate of acceleration is 200 m/sec.sup.2 the time of accelerating the slug is only about half a second for a slug velocity of 100 m/sec at the point of interception and the length of the barrel will only be about 25 meters. The timing sequence for admission of the working fluid from the auxiliary reservoirs C,D,E,F,G,H and I into the chamber during the 6 second cycle is critical and requires accurate timing in the operation of the valves which are preferably under the control of a central computer for controlling the supply of working fluid to the passageways. Thus, the rows of valves C.sub.1 -C.sub.3 associated with the auxiliary reservoir C and which supply the high pressure working fluid spray C.sub.21 -C.sub.23 immediately in front of the barrel opening 54 at the top of the vessel will open in sequence. The first row of valves C.sub.1 will open about 0.1 seconds before the explosion, at 0 plus 2.18 sec. while the last row of valves C.sub.3 will open about 0.115 seconds before the explosion. More specifically, as a slug commences to enter the chamber at time 2.06 seconds before the explosion, the first row of valves C.sub.1 open at 2.08 sec. when expressed in terms of a 6 second cycle, while the last row of valves C.sub.3 open at 2.065 seconds before the explosion. With this time sequence, a slug is allowed to pass out of the barrel opening 54 into the chamber without intercepting the sprays C.sub.21 -C.sub.23 from the rows of passageways C.sub.11 -C.sub.13. Although only three circumferential rows of passageways C.sub.11 -C.sub.13 are shown, it will be understood that a greater number of rows of passageways can be provided in order to obtain the desired amount of working fluid in this area of the vessel. It will also be noted that the row of valves C.sub.3 open before the row of valves C.sub.1 since the distance from the end of the passageways C.sub.13 to the major longitudinal axis of the vessel is greater than the distance from the end of the passageways C.sub.11 to the major central longitudinal axis of the vessel. The rows of valves D.sub.1 -D.sub.4 asociated with the auxiliary reservoir D operate in a similar sequence. Thus, the inner or uppermost row of valves D.sub.1 open at 0.13 seconds before the explosion, preceded by valves D.sub.2 at 0.15 seconds; valves D.sub.3 at 0.18 seconds and, valves D.sub.4 at 0.23 seconds before the explosion. Accordingly, in a 6 second cycle and with the explosion taking place at time 2.18 seconds after commencement of the cycle, the inner row of valves D.sub.1 open at 2.05 seconds while the outer or lowermost row of valves D.sub.4 open at 1.55 seconds after commencement of the cycle. The row of valves E.sub.1 associated with auxiliary reservoir E and connected to the row of passageways E.sub.11 which produce the tubular shaped curtain E.sub.21 of falling working fluid extending from the top to the bottom of the chamber, is opened for the longest duration of time in the 6 second cycle. Thus, the row of valves E.sub.1 are opened at 0.18 seconds in the 6 second cycle or 2 seconds before the explosion and are held open until 0.12 seconds after the explosion which would be time 2.3 seconds in the cycle. The row of valves E.sub.2 would open 1.65 seconds before the explosion and would remain open until 0.9 second before the explosion. The duration of time in which each of the following rows of valves E.sub.3 -E.sub.5 are open is progressively shorter such that the last row of valves E.sub.5 would open at 1.4 seconds before the explosion and closes at 1.2 seconds before the explosion. With this sequence, the successively shorter tubular curtains of working fluid are obtained to form the somewhat spherical pattern of working fluid at the center O of the chamber. The above description is illustrative of one possible method by which a concentrated pattern of working fluid can be obtained near the exploding slugs. Although the working fluid is allowed to flow through the passageways by gravity, this invention can also be practiced whereby the working fluid is placed under a positive pressure such as by the application of a pressurized gas or by the application of a piston to a restricted fluid. The means for carrying out these method are well known and employed, for example, in the field of fuel injection in diesel engines. Of course with such non-gravity injection methods, the spray for the spherically shaped concentration of the working fluid at the center of the vessel could come from all directions of the ellipsoidal vessel. Care must be taken however that the injected spray does not interfere with the flight path or trajectory of the concurring slugs. The wall of the chamber may additionally be protected from the explosions by opening the valves E.sub.2 -E.sub.5, 0.1 second before an explosion and by closing the valves 0.05 seconds after the explosion. Likewise, the valves C.sub.1 -C.sub.3 and D.sub.1 -D.sub.4 remain open until 0.05 seconds after the explosion. Such after-explosion spray from the passageways C.sub.11 -C.sub.13 ; D.sub.11 -D.sub.14 ; and E.sub.11 -E.sub.15 will further diminish the effects from the rising fireball of an explosion. To protect the walls from the shock-blast of an explosion, the valves F.sub.1, G.sub.1 and H.sub.1 are opened 0.5 seconds before and are closed 0.1 seconds after the explosion. The valves connected to the auxiliary reservoir I operate in a similar manner. The row of valves I.sub.1 are opened first and are allowed to remain open for a longer period of time. Specifically, the rows of valves I.sub.1 and I.sub.2 are opened at 0.2 seconds before the explosion and with a velocity of the working fluid of about 3 m/sec. the working fluid is sprayed into the chamber for a distance of about 0.6 meters. The working fluid which is supplied to the row of passageways I.sub.13 at higher pressure is sprayed at a velocity of about 10 m/sec. toward the major longitudinal axis of the chamber and at a time 0.09 seconds before the explosion. The practioner of this invention can easily vary the timing of the working fluid injections and adapt the operation of the system as conditions warrant. By the same token, the size of the vessel wall and the number and sizes of the conduits can also be varied. With the present reactor system, the working fluid is injected into the chamber to obtain maximum attenuation of the explosions and thus maximum protection of the vessel wall, projection 62, and main sliding plates 90. Since the hot working fluid is highly radioactive and concentrated in greater quantities at the bottom of the vessel and especially in the trough 60, a neutron absorbing liquid such as molten lithium is circulated through the chamber wall at the bottom of the vessel. Similar neutron absorbing circulating systems may also be provided in other portions of the wall. Neutron absorbing liquids such as a slurry of depleted uranium in molten sodium or a solution of depleted uranium such as uranyl sulfate or nitrate can also be used. Referring to FIG. 3, it will be noted that the bottom portion of the vessel is provided with a spiral passageway 130 extending through the chamber wall. The neutron absorbing liquid is introduced into the uppermost turn of the spiral passageway through an inlet opening 132 and flows through the spiral passageway 130 from the uppermost loop to a lowermost loop and out of an exit opening 134. The liquid is conducted from the exit opening 134 through a conduit 139 to a heat exchange coil 142 positioned in a heat exchanger 140 and returned through a conduit 138 to the inlet opening 132. A pump 136 is connected in the conduit 138 for continuously circulating the liquid through the spiral passageway and the heat exchanger 140. A heat exchange fluid passing through a secondary heat exchange coil 144 in the heat exchanger 140 is heated and used to perform useful work. Alternatively, the secondary coil 144 can be connected to the conduits 80 in the primary heat exchanger 76. The liquid passing through the coil 142 is thus continuously cooled in the heat exchanger. A drain conduit 150 is connected to the conduit 138 to allow for the periodic draining of the liquid. The drain conduit 150 is provided with a valve 152 to permit drainage of the liquid from the system. A supply conduit 154 is connected to the conduit 138 to allow for the introduction of fresh liquid to the system. A valve 156 is connected in the conduit 154 to allow for the introduction of fresh liquid to the system. If lithium is used as a neutron absorbing liquid, helium and tritium are produced in the liquid. The helium and tritium can be separated from the lithium in a separator facility by methods well known in the art. If a slurry of depleted uranium in sodium is used the separator facility can process the irradiated uranium for fission products and plutonium. Pumping of a slurry presents a greater problem and care must be exercised so that the slurry is maintained fluid enough so that blockage does not develop in the conduits or in the spiral passageway. The fluidity of the slurry can be monitored by sensing devices such as are in everyday use today in the monitoring of chemical slurrys. The purpose of the circulating neutron absorbing and cooling fluid is to absorb neutrons that would otherwise be lost or reflected back into the chamber and to lower the "k" near the wall so that there is a reduction of fissioning within the working fluid near the wall. The advantage of this reduction of fissioning is that a lower breeding ratio is obtained if the actinide fissioning occurs where the neutrons are more moderated prior to fissioning than is the case in the central portion of the chamber. A plurality of systems similar to the neutron absorbing system described above can be provided throughout the wall 12 and also in wall portions in the sump 72; heat exchanger 76 or reservoir 16. These neutron absorbing systems are not only for the purpose of absorbing neutrons and for cooling the walls of the reactor system components but also absorb neutrons so that the quantity of fissioning is greatly reduced. For example if the "k" is 0.70 twice as much fissioning takes place in the working fluid containing actinides as compared to a working fluid having a "k" of 0.40. Since the slurry and precipitate will generate heat from the radioactive decay of fission products and delayed neutrons, it is desirable to lessen the heat producing fissioning in the wall of the vessel, the sump 72, the heat exchanger 76 and the main reservoir 16. Of even greater import is the reduction of fissioning in the slug manufacturing station 100 and in the accelerating mechanism 50. Microscopic actinide particles from exploded slugs are collected and compressed in a mold in the slug manufacturing and assembly station 100 and reused as new slugs in subsequent explosions. Accordingly, the particles which are contained in suspension within the working fluid and which leave through the outlet passageways 70 at the bottom of the vessel can be separated and withdrawn from the working fluid at several locations in the reactor system. Thus, a particles trap 164 is provided in the reservoir 16 and similar traps 165 and 166 are provided in the heat exchanger 76 and in the catch basin 72. The particles entrained in the working fluid settle to the bottom of the traps where they are collected and periodically withdrawn through conduits 168, 170 and 172 connected to the bottom of each trap. A common collecting conduit 174 conducts the particles to the slug manufacturing station 100. A valve 167, 169 and 171 is provided in each conduit 168, 170 and 172 to allow for the withdrawal of such particles at desired time intervals. A common gas outlet conduit 160 is provided for withdrawing residual gases, primarily inert gases entrained in the working fluid, from the top of the heat exchanger 76 and catch basin 72. The conduit 160 is connected to the separator 120 for processing. The vessel 10 and other major components of the reactor system such as the reservoirs, conduits heat exchanger, and the like, must be enclosed within a protective retaining structure, not shown, similar to the structures now employed in conventional nuclear reactors. A pre-stressed spherical containment structure would be preferable for a reactor system using a single pressure vessel since such a structure would lend itself to the separation of safeguard equipment, emergency power supplies, and the like. In a reactor system having a plurality of pressure vessels, a single large cylindrically shaped retaining structure would be preferable. Although the slugs A and B, illustrated in FIG. 1, are shown to be of the same size, this invention can also be practiced where one slug is larger than the other slug. The reactor system can also be adapted so that one of the accelerating mechanisms would propel one of the slugs at a higher velocity than the other slug provided, of course, that the slugs meet near the center of the chamber. This invention can also be practiced wherein a slug is allowed to fall by gravity through the barrel 52 having a slightly larger internal diameter than the external diameter of the slug so that the free falling slug would be guided through the barrel with little friction. In this method the slug accelerating mechanism for the slug is but a long shaft attached to the top of the vessel through which the slug falls until it reaches the chamber. A shaft having a length of about 490 meters would provide a final velocity of about 98 m/sec. and the complex shutter mechanism can be omitted. With liquid sodium as the working fluid, the atmospheric pressure within the chamber 14 will be less than 1 mm. of mercury at the time the slugs enter the chamber so that no serious consideration need be given to the aerodynamic stability or design of the slugs. However, if a slight variation in the flight path of the slugs should occur, even a lateral displacement of the slugs at the point of interception of perhaps 1 mm. would have little effect on the magnitude of the explosion. Similarly, a timing error or a variation in the desired velocity of one of the slugs travelling toward the other slug, such that the slugs would intercept at a distance slightly above or below the intended point of interception, would have little effect on the magnitude of the explosion even if there is a displacement of as much as 15 or 20 cm. from the intended point of interception. The pressure vessel of the present invention is capable of withstanding any variation in the speed of the slugs or in their point of interception within the chamber without causing any damage to the vessel. Of course, if the slugs should fail to meet perhaps by reason of one of the slugs disintegrating before interception, no explosion would take place at all and the debris from the slugs would merely be flushed out of the chamber with the working fluid through the outlet passageways 70. The reactor system would thereafter continue to operate in a normal manner. With a computerized electrical control system, any malfunctioning of the reactor system is highly unlikely since the reliable operation of similar or more complex industrial systems under the control of a computer is well within the capabilities of the present state of the art. In FIG. 4 there is illustrated a detail cross-sectional view of a cylindrically shaped slug 200 for use in the vessel of the present invention. The slug 200 has a diameter of about 30 cm. and a height of about 16 cm. Although the diameter of a slug is more or less fixed by the mechanical construction of the accelerating mechanism 50 or the size of the barrel 52 through which a slug is propelled, the height and the volume of the fissile-fertile material in the slug may be varied if it is desired to increase or decrease the yield or the amount of energy released by an explosion. The concentrations of different actinides within the fissile-fertile material can also be varied or the velocity of the slugs can be increased or decreased if it is desired to increase or decrease the magnitude of an explosion. Of course, the concentration of actinides within the fissile-fertile material must be maintained within the safety limits of the vessel and support equipment so as to prolong the life of this equipment over as long a period of time as possible. Each slug 200 is formed of a major cylindrical component 204 of a metal or metal alloy. If sodium is used as the molten metal working fluid than the cylindrical component is made of sodium. The sodium should be substantially free of insoluble materials in suspension and of dissolved metals such as rubidium or cesium. Sodium of a greater purity could be obtained by distillation but such processing would be more expensive and would therefore not be as economical. The primary function of the cylindrical portion 204 is to reflect fast neutrons back into the fissioning fissile-fertile material and to act somewhat as a tamper on the explosion. The forward or frontal portion of the cylindrical slug 200 is provided with a hemispherical portion 202 having a radius of about 12 cm. The hemispherical portion is formed of a compressed material which was collected as a precipitate from the working fluid containing the debris from previously exploded slugs. The composition of the material in the hemispherical portion 202 is about 70% uranium, 28% plutonium and the balance of 3% comprising other actinides, fission products, sodium, and a trace of corroded material from the wall of the vessel or from the conduits, heat exchanger, reservoirs or catch basin. The hemispherical portion 202 is positioned centrally of the cylindrical portion 204 so that its flat face 206 is flush with a flat frontal surface 208 on the cylindrical portion 204. The manufacture of the slugs in the slug manufacturing station 100 is performed automatically and by remote control since the slug materials are highly radioactive. Thus, the manufacture of the slugs can be performed in three basic steps, as follows: 1. Forming of the cylindrical alkali metal portion 204 by conventional molding or casting techniques. The cylindrical portion 204 is provided with a hemispherical recession or indentation in the frontal surface 208 corresponding in size to the hemispherical portion 202; PA1 2. Forming the hemispherical portion 202 by pressing the collected precipitate in a mold such as is conventionally done with metal powders, and PA1 3. Inserting the hemispherical portion 202 into the indentation in the cylindrical portion 204. The metal of the cylindrical portion is maintained close to the melting point so that the metal will be soft enough to cause bonding between the metal of the cylindrical portion 204 and the hemispherical metal portion 202. PA1 1. The heat flash from an explosion is attenuated by the metal working fluid which is mixed with the debris from previous explosions so that substantially none of the heat flash produced by the explosion will reach the wall of the vessel. PA1 2. The shock-blast of the explosion is, to a great extent, attenuated by the inertia of the working fluid which is injected into the chamber and by the lack of an atmosphere within the chamber. PA1 3. Neutrons are absorbed by the working fluid itself and by the debris in the working fluid to moderate and absorb neutrons. On the average the fission-born neutrons will need to make more than 200 collisions with the atoms within the chamber before their energy level is low enough so that capture by sodium becomes probable. With about 10,000 atoms of sodium, there is about 12 atoms of U.sup.238, about 5 atoms of plutonium, and a balance of about 3 atoms of magnesium from neutron irradiated sodium, wall materials, fission products, or other actinide isotopes. Most of the near-thermalized neutrons will be captured by U.sup.238 which will eventually decay into Pu.sup.239. More neutrons will fission plutonium than will be captured by plutonium and still fewer neutrons will be captured by sodium or other isotopes. Few neutrons will reach the wall of the vessel and most of those neutrons which do pass through the working fluid to irradiate the wall are delayed neutrons or neutrons from fissioning that was caused by delayed neutrons. Over an extended period of time, the wall will become radioactive, but with a careful selection of materials which are resistant to neutron irradiation, damage to the wall will be negligable. PA1 As has been previously disclosed a plurality of neutron absorbing systems are provided within the wall. Accordingly, few neutrons will be absorbed by the wall itself and less fissioning will occur near the wall so that fewer neutrons will cause neutron irradiation of the wall. PA1 4. Most of the slug material and of the nearby working fluid is "plasmatized" (a majority of the atoms are ions). These, more active chemical ions, are "neutralized" by the injection of an additional quantity of working fluid into the chamber so that there is a minimal corrosion of the wall by such ions. In addition to acting as a neutron reflector and as a tamper, the metal of the cylindrical portion 204 provides a lubricating medium by which a slug is more readily able to pass through a barrel 52. Since metals such as lead, sodium, lithium, or a sodium-lithium alloy, are of a soft and non-abrasive composition, no damage will be done to the barrel as the slugs are propelled into the chamber. The slugs are formed of a size so that they "fit" into the barrel so that as little as possible of the propelling helium gas will escape between the slug and the barrel yet without retarding movement of the slug through the barrel. Another embodiment of a slug 210 is illustrated in FIG. 5. In this embodiment, the rear portion of the slug need not be provided with a flat surface. Accordingly, the slug has a part-cylindrical configuration in which a cylindrical portion 211 of about 4 cm. in length is provided as a guide surface for the slug as it is propelled through the barrel. The remaining rear portion 213 of the slug is part-hemispherically in shape having a radius of about 15.5 cm. A hemispherically shaped portion 212 having a radius of about 12 cm., formed from a compressed material of the same composition as the portion 202 in the slug of FIG. 4 is positioned centrally of the cylindrical slug portion 211 such that the flat front face 218 of the hemispherical portion 212 is flush with a flat front face 220 on the part-cylindrical portion 214 to thereby form a combined flat surface. The slug 210 is constructed in the slug manufacturing station 100 in substantially the same manner as the slug 200 of FIG. 4. The following illustrations are presented for help in understanding the broad general principles and it is recognized that in actual practice, different values can be applied. If each of the concurring slugs has a velocity of 200 m/sec. the slugs are approaching each other at about 0.4 mm per micro-second. Since there are about 11 kg of Pu-240 within the combining slugs and about 1,000 neutrons per second are produced from each gram of Pu-240 by spontaneous fissioning, and there are neutrons from the spontaneous fissioning of other actinides roughly 12 neutrons will be produced every micro-second. If the prompt "k" of the combining slugs is 0.99 there is a multiplying effect of about a hundred times. Accordingly, as the slugs approach each other much fissioning takes place and many neutrons are being produced even though the combining slugs are not prompt critical. Very rough and simplified calculations show that at the instant the slugs become prompt critical there are about 10.sup.4 fissionings per "generation" of about 20 nano-seconds. After about 15 micro-seconds the prompt "k" is say 1.07 and over 10.sup.16 fissionings will then be occurring per "generation". At such time the energy from such fissioning will cause the slugs to begin to vaporize and explode. Then the "k" begins to decline even though the slugs are still concurring. Even so the rate of fissioning continues to rise for a couple more micro-seconds. As the combining slugs are dissipated or exploded, the "k" falls below 1.00 and the rate of fissioning rapidly declines. Because of delayed neutrons, fissioning will continue even though the debris from the exploded slugs is now entrained within the working fluid in the chamber. Say that about 10 milli-seconds after the explosion in the centroidal 100 cubic meters of the chamber the "k" is about 0.5 and that means that the neutrons will be doubled within that centroidal portion. The reactor system is designed to contain an explosion with an energy burst of about 4.times.10.sup.10 joules in say a milli-second and have an additional 8.times.10.sup.9 joules of energy from energy of delayed neutron fissioning and radio-active decays. All this energy comes directly or indirectly from the fission of about 0.6 grams of actinides which is about 1.5.times.10.sup.21 fissionings. Less than four actinide atoms of a million atoms of the combined slugs fission. It will also be readily apparent from the preceding description of the slugs that they are most unsuitable as components in atomic weapons. Present day atomic explosives have an efficiency (fraction of actinide atoms fissioned) that is perhaps 10,000 times greater than the slugs of this invention. Although the fission explosions obtained with the slugs of the present invention are of an extremely low yield, it is so intended for reasons to be explained subsequently. If the slugs are propelled to attain a velocity which is the minimum of the normal or intended velocity for the slugs, the energy yield per cycle would be approximately 4.8.times.10.sup.10 joules. However, if the velocity is below the minimum velocity, the combining slugs will be blown apart or dispersed before there has been enough fissioning to produce the desired quantity of energy. An explosion producing a lessor magnitude of energy presents no great loss because an adjustment in the speed of the slugs or a slight decrease in the time intervals between explosions would readily make up for the loss of energy not produced by such a "fizzle" explosion. Although a "fizzle" explosion would result in a slight increase in the costs of fabricating and accelerating an extra set of slugs, it does not represent a loss in the amount of fissle actinide materials available in the system since the debris from the "fizzle" explosion would be extracted from the working fluid and re-assembled in the manufacture of additional slugs in the slug assembly station 100. The total quantity of fissile actinide materials within the very lean sodium slurry working fluid, that is within the chamber 14 at the instant of an explosion, will be hundreds of kilograms and with about 140 metric tons of such working fluid within the chamber the "k" could be about 0.30. If the "k" were to be say 0.65 there would be about twice as much fissioning therein as with a low "k" of but 0.30. Thusly, control of the quantity of the debris in the working fluid can be used to control the magnitude of the explosions in the chamber. The ratio of the fissile actinide material, i.e., plutonium, can readily be varied with respect to the fertile material, i.e. uranium, so that the "k" of the debris in the working fluid and in the slugs is arranged such that the desired magnitude of explosion is possible. With the various factors which lend themselves to ready change, such as the speed and composition make-up of the slugs, composition and quantities of actinides in the slurry, and of the total actinide mass of the slugs, explosions producing the desired quantity of total energy can be obtained. Likewise, explosions of greater than desired magnitude can easily be avoided. An advantage of the present reactor system over the liquid sodium cooled, fast neutron reactors now in use is that no cladding material which moderates and absorbs neutrons, is needed. Much of the fissioning which takes place in the explosion of slugs in the present invention occurs without any of the neutron absorbing and neutron moderating working fluid being present. Only after most of the fissioning has taken place, do the actinide materials mix with the working fluid to form a slurry. There are no oxygen atoms to moderate the fission-born neutrons. It is expected that over 20% of the fissioning will be the fast neutron fissioning of Uranium-238. This is higher than what can be obtained with a sodium cooled breeder reactor. There will be more neutrons from each fissioning because there are more neutrons per fission from fast neutron fissioning than from slower neutron fissioning. For example, thermal neutron fissioning of Pu-239 yields an average of 2.89 neutrons per fission, 1.3 Mev neutrons yield 3.08 neutrons, and 4.0 Mev neutrons yield 3.43 neutrons per fission. Also with fast neutrons there tends to be a higher ratio of fissionings to capture and this is particularly true with such neutron absorbing isotopes as Np.sup.237 ; Pu.sup.240 and Pu.sup.242 all of which are fissionable by fast neutrons. With a concentration of debris in the working fluid having a higher "k", the energy produced by delayed neutron caused fissioning will be significant. Fissioning Pu.sup.239 has 0.21% delayed neutrons while fissioning U.sup.238 has 1.57% delayed neutrons. For example, I.sup.137, which has a half life of 22 seconds and which decays by neutron emission, produces 0.215% of the neutrons from U.sup.238 fissioning and 0.063% of the neutrons from Pu.sup.239 fissioning. By far the most important isotope breeding reaction of this embodiment is the following series of nuclear transformations: ##STR1## Accordingly, fuel for the manufacture of additional slugs for use in the reactor system of the present invention is bred. Since the system produces an excess in fuel, such additional fuel can be used either in conventional reactors as a "plutonium re-cycle" or in other new reactors of this invention. In a further modification, the composition of a slug could contain actinides which are undesirable in nuclear reactors in use today. For example, Np.sup.237 is produced by the faster neutron stripping of an extra neutron from U.sup.238 and the subsequent decay of U.sup.237. Thus, in present day reactors Np.sup.237 will capture a neutron and Np.sup.238 will decay into Pu.sup.238 which will most likely capture another neutron before it fissions to produce Pu.sup.239. However, with the very fast neutrons produced in the explosions of the reactor of the present invention, it is expected that Np.sup.237 will fission rather than become Np.sup.238. Moroever, in the explosion itself there is the likelihood that Np.sup.238, with a half-life of 2.1 days, will fission with its very high fission cross-section. Likewise, Pu.sup.238 is more likely to fission rather than to absorb a neutron. In other words, Neptunium can be used as a fissile material in the slugs of the present invention. As a further alternative, thorium can be used as fertile material in the slugs of the present invention. Although thorium is less likely to fission to fast neutrons than is U.sup.238, thorium will react and decay when irradiated by neutrons, in accordance with the following reactions: ##STR2## After the reactor has been in operation for a time and after a predetermined number of explosions, the actinides are processed and protoactinium is isolated by methods not specifically described in the present application since they are well known in the art. The protoactinium decays, after several months, to form uranium which will be rich in the most fissile isotope U.sup.233. In other words, the materials making up the hemispherical portions 202 and 212 can be designed to contain plutonium, thorium and uranium containing isotopes 232, 233, 234 and 238. The protoactinium produced in the explosion when thorium containing slugs are used can be isolated and removed by methods well known in the art. Although there are many possible variations in which thorium can be used as a fertile material to produce U.sup.233, the system described in the preceding paragraph is preferred because high-purity (uranium)u 233 is obtained and more fast fissioning of fertile isotopes is produced than would be the case if only fertile thorium were used. In other words, the net input in this (thorium) embodiment would be 44 thorium atoms and 82 depleted uranium atoms and the net output would be the fission products from 100 fissionings, 1 atom of (protactinium)Pa-231, and 25 atoms of U.sup.233. Note that most of the fissioning is of plutonium which was bred from depleted uranium. The easily isolatable U.sup.233 is much more valuable for slow neutron reactors than is plutonium. In order to operate the system of the present invention as economically as possible, there will be less processing of the working fluid and precipitate for fission products and hence a larger build-up of fission products in both the working fluid and the re-processed slugs. This will, in effect, reduce somewhat the breeding ratio, but oalculations indicate that for every 100 fissions of actinides there will be produced about 16 atoms of Pu.sup.239 ; 7 atoms of Pu.sup.240 ; 4 atoms of Pu.sup.241, 2 atoms of Pu.sup.242, and one other atom of some other actinide isotope in excess of those needed in the operation of the reactor system. One hundred and thirty depleted uranium atoms will be the net input in this illustration of the preferred or "depleted uranium" embodiment of this invention. In the practice of the preferred embodiment of this invention the exact mass of actinides, the fractions of fissile and fertile actinides therein, and the "k" of the slurry should so be adjusted that the two concurring slugs near the center of the chamber 14 when substantially surrounded by spray-slurry become a more than prompt-critical assembly when their front surfaces are about one centimeter from each other. If the single slug after manufacture is tested in the sub-critical mode, by means known to those skilled is such arts, and is found to be too low in fissionability for the above conditions to be met, it could be programmed to be paired-explosionated with a slug that is of higher fissionability but not close to prompt-criticality by itself. The fissionability of a slug can most easily be adjusted by changing the percentage of plutonium in the portions 202 or 212. Also the size of the portions 202 or 212 can be changed. For example, the radius r could be as small as say 11 cm or as large as say 13 cm without necessitating any change in the sizes of the slug accelerating means. As there is a build-up of fission products and an increase in the percentage of fissile actinides therein, such changes would be needed. Note that to decrease the fissionability of slugs with a fixed size portions 202 or 212 some more uranium could be added to the materials used to construct such portions. The uranium used for the practice of this invention might be depleted uranium which is about 0.2% uranium-235 and over 99.7% uranium-238. Such depleted uranium is in great over-supply and is priced at about 3% of the price of natural uranium. Also, uranium from spent enriched light water reactor fuel assemblies could be used. Such uranium averages about 0.8% uranium-235, about 0.2+% uranium-236, and about 98.9+% uranium-238. It should be noted that with such uranium there would be more fissioning of uranium-235 and there would be more production of neptunium-237 that with depleted uranium. It should be noted that uranium-236 has larger cross-sections for fissioning and fissions to slower neutrons than does uranium-238. Perhaps the best source of plutonium for use in the practice of this invention would be the plutonium from spent fuel of "plutonium re-cycle" where the plutonium isotopes Pu-238, Pu240 and Pu-242 are of much higher percentages. For the best breeding of isotopes or for less wear on the reactor it is desired to have a minimal fraction of actinides in the lean sodium slurry which is the working fluid. From the figures shown hereafter the atomic percent of actinides within the slurry is about 0.20% while the weight percent of actinides is about 2.07%. Such a slurry is harder to pump and is more corrosive and/or erosive upon the passageway wall, etc. than is purer molten sodium. Also, the larger the quantities of actinides within the slurry the greater the fractional proportion of fissionings within the slurry. A larger fraction of fissionings with the slurry will result in a lower breeding ratio and in greater undesireable irradiation of the chamber 14's walls by both neutrons and gamma rays. However, the use of a leaner lean slurry would require more time for precipitations and a larger inventory of actinides and sodium. Hence there would be a longer doubling time and greater tool expense for the construction of larger storage basins 16, etc. Also the slugs would have to be made slightly more fissionable. Although the plutonium can be withdrawn and used in other types of fast neutron reactors or in a reactor using a "plutonium re-cycle" the best use of the material bred in the reactor system of the present invention is to use the fissile material in another reactor system of this invention. With rapid precipitation and fabrication of slugs a "doubling time" of but a couple of years can be obtained. Such quick processing means that some delayed neutrons will be emitted within the fabricated slug. For example, nine minutes after fissioning from a hundred million fissionings, about 22 delayed neutrons will still be "born". The effects of a nuclear explosion within the chamber 12 may be summarized as follows: The reactor system of this invention can be used for the prdocution of isotopes useful in fusion reactions as well as the production of fissile isotopes. For example, if lithium is used as the working fluid, tritium will be produced from the reaction: EQU n+Li.sup.6 =T+He.sup.4 +4.8 Mev. (3) Tritium is radioactive and with a half-life of 12 years decays into Helium-3. Both tritium and He.sup.3 are usable in fusion reactions. Natural lithium contains 7.4% lithium-6 which has a thermal neutron cross-section for the Li.sup.6 (n,T)He.sup.4 reaction of 950 barns and contains 92.6% lithium-7 whose thermal cross-section for the Li.sup.7 (n,T)Li.sup.8 is a very low 0.036 barns. Lithium-8 has a half-life of 0.85 seconds and decays into beryllium-8 which, after a half life of 10.sup.-16 seconds, fissions into two helium-4 atoms. Lithium as a working fluid has advantages over sodium in that it has a much lower vapor pressure and a much higher specific heat. The heat of vaporization for lithium is more than 4700 calories per gram. However, lithium is more of a moderator and to some metals is more corrosive than sodium. If lithium-7 metal (depleted lithium) is used, it would have a few reactions to very fast neutrons: EQU n+Li.sup.7 .fwdarw.Li.sup.6 +2n (slower) (4) Li.sup.6 from reaction (4) would produce tritium and helium by the reaction (3). With lithium-7 as the working fluid very little tritium would be produced and that tritium could remain within the working fluid as lithium (tritide). Over a period of time, tritium will decay into helium-3 which is easily separated and collected as a gas. Helium-3 with its very high cross-section to thermal neutrons of 5500 barns can find use as a very rapidly moving reactor control gas. (At 2352.degree. K., He.sup.3 has an average velocity of about 7000 meters per second). Also it can be used as a fuel for the fusion reaction: EQU D+He.sup.3 .fwdarw.H+He.sup.4 +18.3 Mev (5) Both of the products are as ions and with certain types of conversion systems would be much desired over fusion reactions that produce neutrons. If natural lithium is used, great quantities of tritium would be produced in the reactor system of the present invention. Those skilled in the art can find much discussion in current literature as to how tritium can be separated from molten lithium. The use of lithium as the working fluid would lower the "k" of the actinides in slurry with the lithium and hence there would be a need to increase the percentage of fissile material in the slugs if the invention is to be practiced with the same parameters as with sodium. Taking into account the high cost of lithium, however, the best means of producing tritium in the reaction system of this invention is to use an alloy of about 5% lithium in sodium. With such a molten alloy, for every thermal neutron captured by sodium there would be about 7.4 atoms of tritium produced. The pressure vessel of the present invention can be constructed by methods which are well within the present state of the art. More specifically, the vessel could be constructed by erecting an inner "skin" of a heavy gauge steel having a thickness of about 5 to 10 cm. The skin forming the inner surface of the wall 12 could be erected by welding together a plurality of plates and by holding the plates in position by means of an interior and exterior frame. Components of the frame could be introduced to the interior of the skin though the barrel openings 54 which have a diameter of 30 cm. or through the passageways 70 which have a diameter of 40 cm. if 3 such passageways are provided. If these openings should still be too small, a larger access opening can be provided in a convenient place in the inner skin. The plates forming the inner skin are provided with a plurality of openings and a conduit of predetermined length would be attached to each opening to thereby form the passageways extending through the wall of the vessel. Additional conduits of predetermined length would be attached to the inner skin to form the outlet passageways 70 and the barrel openings 54. An outer skin of heavy gauge steel plates, forming the outer surface of the wall 12 and having a corresponding number of openings would be erected and fitted to the conduits forming the passageways to thus form a skeleton vessel having inner and outer skins with all of the conduits which form the passageways providing effective supporting struts to hold the outer skin in position relative to the inner skin. Metal pellets or coarse granular metal particles could be poured into the space between the skins followed by the introduction of a quantity of molten metal which will flow into the spaces between the pellets or grain to form a unitary and solid wall portion. The process can be carried out by a continuous casting process in which a layer of pellets or grain is introduced into the space between the skins followed by a stream of molten metal. The process is continued until the entire space between the skins is filled to form the ellipsoidal pressure vessel. The hot molten metal when coming into contact with the inner and outer skins and the pellets or metal particles produces a welding or joining of the metals with the molten metal to thus weld the skins and pellets or grains to produce a unitary structure or wall 12. Alternatively, molten metal can be poured into the space between the skins without the use of metal pellets or grain. Suitable supporting means such as struts support the vessel within a spherical or cylindrical containment structure which would be constructed of a thick wall of reinforced concrete in a manner similar to the containment structures which are used in present day nuclear reactors. The reservoirs may be supported upon ledges provided on the interior wall of the retaining structure or they may be supported by means of struts or supporting beams mounted on the interior wall of the supporting structure or, if space permits, on the exterior of the pressure vessel wall 12. If a plurality of vessels are erected in a common "Nuclear Reactor Park", a common slug manufacturing station, a common debris purification system, and a common outer containment vessel can be provided. The pumps, heat exchangers, and working fluid can be arranged so that the system can be used even if one of the chambers is closed down for maintainence or repair. It is believed that the present invention comprises a solution to the serious problems of energy shortages existing in major parts of the world today. Moreover, the present invention provides an alternative method of producing thermal energy and of providing valuable isotopes for the nuclear breeder reactors which are now under development. The nuclear explosion breeder reactor system of the present invention can be constructed by methods which are well within the state of present day technology. The reactor system is economical to operate and capable of producing energy without depleting present day resources of enriched uranium. Finally, the reactor system of the present invention is designed to breed valuable isotopes. This invention could be practiced with larger contained explosions, with different sizes of slugs, with leaner slurry, and with a lower fraction of fissile actinides. For example, the velocity of the slugs could be such that an explosion equal to about 2.8.times.10.sup.11 joules about every 7 seconds yields about 48 billion watts of thermal power. The slug could have a diameter of say 40 cm. and the radius r of the portions like 202 or 212 could be say 15 cm. That means that the portions like 204 or 214 would have greater tamping and neutron reflecting functions. The plutonium within the slurry and in the slugs will be about 20% of the actinides while the fraction of actinides within the slurry will be say one-half of the fraction in the preferred embodiment. Likewise, the quantities of spray and the sizes of the chamber, pumps, slug accelerating means, etc. would also be scaled upwardly. Note that other accelerating means such as the magneticly driven "mass driver" could be used. Of course these and other modifications could be made in this invention without deviating from this invention's broad concepts. It is understood that the foregoing description is illustrative of a preferred embodiment of this invention and that the scope of the invention is not to be limited thereto but is to be determined by the appended claims. |
048329045 | summary | FIELD OF THE INVENTION The invention relates to an emergency cooling device for a fast neutron nuclear reactor of integrated type. BACKGROUND OF THE INVENTION Fast neutron nuclear reactors of integrated type comprise a main vessel containing liquid metal such as sodium, forming the liquid for cooling the nuclear reactor in which is immersed the reactor core consisting of fuel assemblies. The main vessel of the reactor is divided internally into two regions, by means of a complex structure forming the inner reactor vessel. This complex structure is equivalent to a wall, a part of which, known as a stepped wall, extends radially relative to the main vessel. Of the two regions bounded by the inner vessel, the first is arranged substantially in the upper part of the vessel and the other in the lower part. The upper region, known as the hot header, communicates with the core outlet and receives the hot liquid metal which has passed through the core fuel assemblies. The lower region, known as the cold header, receives the sodium cooled in the intermediate exchangers immersed in the main reactor vessel. This cooled liquid metal is then conveyed from the cold header to the lower part of the core assemblies. When a nuclear reactor has operated for a certain period of time, it continues to release significant residual power when it is stopped, i.e. when the reactor control rods are inserted into the core, in their maximum insertion position. The residual power of the reactor must therefore be removed to avoid damage to internal components and structures owing to an excessive temperature rise. This possibility of removing the residual reactor power must be maintained even when the reactor has undergone major breakdowns and when the main power removal circuits, which are employed when the reactor runs normally, are out of action. Emergency circuits are therefore resorted to, these being used only when the reactor is stopped and when the main circuits are out of action. In the case of fast neutron nuclear reactors cooled by liquid sodium and of an integrated type, use is made of emergency heat exchangers immersed in the reactor vessel, inside the hot header. These emergency heat exchangers, in which the liquid sodium reactor coolant circulates, are associated with heat exchangers of the sodium-air type, arranged outside the reactor vessel and responsible for cooling the secondary liquid sodium which has been heated by coming into thermal contact with the primary sodium reactor coolant, passing through the exchanger immersed in the vessel. These emergency exchangers, associated with sodium-air exchangers, form circuits which are completely independent of the main circuits. The emergency exchangers, which are immersed directly in the hot header of the vessel, comprise inlet openings for the sodium to be cooled in their upper part and outlet openings for the cooled sodium in their lower part. The cooled sodium is therefore reintroduced into the hot header and must follow a complicated path in order to travel into the cold header and, from there, to return to the lower part, or bottom, of the fuel assemblies. This complex path includes passing through pumps and the bodies of the intermediate exchangers of the main power removal circuits. This results in fairly large pressure drops, reduced efficiency of the reactor emergency cooling device and considerable temperature dissymmetries in the hot header, leading to additional thermal stresses on the stepped wall. SUMMARY OF THE INVENTION The objective of the invention is therefore to provide an emergency cooling device for a fast neutron nuclear reactor of integrated type having a main vessel containing liquid metal in which the reactor core is immersed and comprising a wall dividing the inner space of the vessel into an upper region receiving the hot liquid metal which has passed through the reactor core, called the hot header, and a lower region receiving cooled liquid metal, called the cold header, this wall comprising a part extending in the radial direction from the vessel, called the stepped wall, and the emergency cooling device comprising at least one heat exchanger immersed in the hot header and having outlet openings for the cooled liquid metal in its lower part, the device being highly efficient and very simple to use. To this end, the cooling device according to the invention also comprises: a substantially vertical tubular conduit passing through the stepped wall vertically below the exchanger, whose lower end opens into the cold header and whose upper end opens into the hot header, below the lower part of the exchanger; a bell fixed onto the heat exchanger, in the extension of its lower part, open downwards and arranged so as to cap the upper part of the conduit while providing a free space around this upper part; and a means for compressing inert gas and for relieving pressure in the inner space of the bell enabling the liquid metal in the hot header to be completely separated from the liquid metal in the cold header, by means of compressed inert gas, or of bringing the outlet for the cooled liquid metal from the heat exchanger into communication with the cold header, by means of the bell and the conduit. |
abstract | A method of collecting 3He from a nuclear reactor may include the steps of a) providing heavy water at least part of which is exposed to a neutron flux of the reactor, b) providing a cover gas in fluid communication with the heavy water, c) operating the nuclear reactor whereby thermal neutron activation of deuterium in the heavy water produces tritium (3H) and at least some of the tritium produces 3He gas by β− decay and at least a portion of the 3He gas escapes from the heavy water and mixes with the cover gas, d) extracting an outlet gas stream, the outlet gas stream comprising a mixture of the cover gas and the 3He gas and e) separating the 3He gas from the outlet gas stream. |
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058833943 | summary | A portion of the disclosure of this patent document contains material which is subject to copyright protection. The owner has no objection to the facsimile reproduction by anyone of the patent document or the patent disclosure, as it appears in the Patent and Trademark Office patent file or records, but otherwise reserves all copyright rights whatsoever. This application claims priority from U.S. Provisional Application Ser. No. 60/008,357, filed Dec. 07, 1995, the disclosure of which is incorporated herein by this reference. TECHNICAL FIELD OF THE INVENTION This invention relates to novel, improved devices for radiation shielding, and to methods for fabricating and using the same. Devices of that character are well suited for use in protecting personnel from ionizing radiation in nuclear power plants, and in particular in protecting personnel from receiving doses of ionizing radiation which are emitted from structures such as pipe and valves. BACKGROUND OF THE INVENTION A continuing demand exists for a simple, inexpensive radiation shield which can be used to simplify the installation and/or removal of shielding material as may be necessary to achieve reductions in dose exposure to personnel in nuclear power operations, particularly as regards exposure to gamma radiation. This is true today, even though flexible lead shielding, or encased lead shielding has heretofore been widely used for providing certain types of radiation shielding. In fact, in older nuclear power plants, the dose exposures to plant workers due to accumulations of radioactive materials in plant structures, and in particular, pipe runs, increasingly have brought this problem to the attention of plant operational personnel. During maintenance and overhaul of power plant systems, personnel are frequently required to perform operations that bring them into close proximity to locations which have the potential to accumulate, and thus emit, potentially harmful ionizing radiation. Generally, the prior art apparatus and methods known to me are too cumbersome, and they are not particularly well adapted to being secured in place for long term radiation protection. As a result, the overall radiation dosage received by nuclear plant workers could be appreciably reduced with the availability of improved radiation shielding devices, and in particular, with improved devices that are suitable for being left in place for continued use during long term plant operations. It would also be desirable for a number of reasons to be able to utilize multiple radiation shield portions in a radiation shield application. First, multiple shield portions could be used to increase the shielding effect, by combining the shielding capability of multiple layers of radiation shield portions. Second, multiple shield portions could be utilized to efficiently accommodate a varying dimensional requirement, such as the curve of a pipe, or an elbow in a pipe run. Third, many fabrication personnel find that it would be desirable to have radiation shield portions which can be conveniently fastened together to produce a final radiation shield of a desired size, in an easy, building block fashion. One important problem which must also be overcome with respect to any lead based radiation shield design is the potential for contamination of lead by existing radioactively contaminated materials, as that would result in further contamination since the lead may itself become radioactive. In other words, the use of a lead shield necessitates protection of such a lead shield, to avoid the possibility of further contamination, of either the lead itself, or of the underlying area due to lead becoming deposited thereon. This problem is further aggravated when the shields are placed in locations subject to high temperature or to water spray. Depending upon the anticipated service, a radiation shield may be subject to various adverse or harsh operating conditions, and thus the design must accordingly be capable of reliably protecting the lead during such service. Currently, when it becomes necessary to work on or near pipe runs which are emitting an appreciable radiation dosage, common practice has been to use a type of wool blanket, or lead shot bags, or lead strips. Each of such apparatus and the methods for their use are somewhat effective in reducing radiation dosage, but in each case, their use has certain drawbacks, including: (1) the equipment is too bulky (especially in the case of a lead wool blanket); PA1 (2) the equipment is prone to leak (such as in the case of lead shot bags, where loss of lead causes other contamination problems); and PA1 (3) installation of the apparatus is too time consuming (such as in the case of installation of lead sheet strips). PA1 can be used in radioactively contaminated areas with minimal risk of contamination by the lead from the shield; PA1 can be provided in a simple coating that allows use in moderately moist environments; PA1 can be used where the shielding is not expected to encounter high temperatures; PA1 can be used where the shielding is not expected to encounter high pressure water spray; PA1 which can be used in direct contact with stainless steel piping and components; PA1 are relatively simple, particularly in the manufacture and installation, to thereby enable the devices to be easily prefabricated and installed for unique applications; and PA1 which can be easily decontaminated. PA1 can be easily used with stainless steel plate as the encapsulating material, so as to allow use in areas which may encounter high pressure spray; PA1 can be used in radioactively contaminated areas with an absolute minimum of risk of contaminating the lead in the shield; PA1 can be used on or around piping and components requiring that the shielding be protected against moisture, heat, and high temperature water or steam; PA1 can be left inside the primary containment building during operation of the nuclear reactor plant. PA1 compatible with direct stainless steel contact; PA1 easy to decontaminate; PA1 able to withstand short duration exposure to water or spray; PA1 able to withstand short duration moist temperatures to about 145.degree. F.; and PA1 are able to withstand moderate flexing and bending, without cracking and peeling. PA1 compatible with use in moderately high temperature environments (up to 450.degree. to 500.degree. F.); PA1 able to withstand prolonged exposure to moisture and high pressure water or steam spray; PA1 are able to withstand moderate flexing and bending; PA1 easy to install and to remove. Although at least one proprietor has recognized the need for an improved radiation shielding that is available in sizes that can be manipulated by hand by a single workman, and which protects the underlying structure from lead contamination, unfortunately, such devices known to me have left something to be desired. Consequently, I have developed novel radiation shielding designs, and methods for their fabrication, and for their installation, which provide radiation shielding which is superior to earlier radiation shielding apparatus and techniques which are known to me. Radiation shielding devices which provide some of the general capabilities desired have heretofore been proposed. Those of which I am aware include those described in the following patents: U.S. Pat. No. 5,012,114 issued to Sisson on Apr. 30, 1991 for a Radiation Shield; and U.S. Pat. No. 3,785,925 issued to Jones on Jan. 15, 1974 for a Portable Radiation Shield for Nuclear Reactor Installation. The patent documents identified in the preceding paragraph disclose devices which do not provide permanently affixable radiation shield designs, and thus are inherently not as well suited, as disclosed, for many of the applications which are of interest to me. The radiation shielding devices provided by Sisson are not suitable for exposure to moderate or high temperatures, or to water spray environments, due to use of a vinyl plastic sheet as a protective surface material. And, the portable shield provided by Jones, which is designed for protection of the dry well during removal of fuel from a BWR plant, though it involves the provision of a lead filled stainless steel shielding device, is so large and unique as to be inapplicable for most of the smaller applications of interest to me. Therefore, there still remains an unmet and increasingly important need in the field for a radiation shield which is designed and manufactured in a way that assures sufficient structural strength to withstand use for either permanent or temporary service in harsh conditions, and which have the assurance that retrieval is possible without encountering adverse lead contamination. Thus, the advantages offered by my novel radiation shield designs, which are permanently mountable (even in highly controlled locations such as a dry well) and which may be provided in sizes which are transportable by a single worker, yet be removable and cleanable, are important and self-evident. SUMMARY OF THE INVENTION I have now invented, and disclose herein, a novel, radiation shield for use in attenuating exposures of radiation workers to ionizing radiation. Unlike radiation shields heretofore available, my shields are simple to build, particularly for custom applications, easy to install, relatively inexpensive, easy to use while avoiding undesirable lead contamination, and are otherwise superior to the heretofore used or proposed radiation shield devices of which I am aware. In one exemplary embodiment my radiation shield is provided in a sequence of layers of radiation shield portions, wherein the sequence of layers of radiation shield portions comprises a first shield layer S.sub.1 through an Nth shield layer S.sub.N, wherein N is an integral number corresponding to the number of radiation shield layers provided. In each of the shield layers S.sub.1 through S.sub.N, a one or more, and preferably a plurality shield portions is provided. The number of shield portions in each layer may be described, in sequence, as shield portions S.sub.1 (1) through S.sub.1 (X) occurring in a first shield layer, through shield portions S.sub.N (1) through S.sub.N (X) in the Nth shield layer. For each layer, a positive integer X, one or larger, describes the number of shield portions in that layer. In one embodiment, my radiation shields are provided with flexible, elastomeric coating, with a minimum of 60 percent elongation, and a Shore D hardness of about 37, and with a tensile sheer strength of approximately 347 pounds per square inch. The coating is preferably provided in a Bisphenol A epoxy which is cross-linked with a modified cycloaliphatic amine curing agent. Grommets are provided for ease of hanging the shields. The shields can be provided in the shape of a segmented annulus, up to half-round form, or in planar sheet form. In another embodiment, my shields are provided with an inner layer of at least one sheet of solid radiation shielding material, and an outer stainless steel casing. A sealant, preferably silicon sealer type, is used between at least portions of the solid radiation shielding material and the outer stainless steel casing. The sealant and the stainless steel casing cooperate to effectively seal the solid radiation shielding material against leakage outward through the outer stainless steel casing. My novel radiation shields are simple, durable, and relatively inexpensive to manufacture. In use, they provide a significant measure of reduction in radiation exposure to workers, by virtue of their ease of use in areas which were heretofore difficult to shield, and thus provide a significant improvement a radiation shield device. OBJECTS, ADVANTAGES, AND FEATURES OF THE INVENTION From the foregoing, it will be apparent to the reader that one important and primary object of the present invention resides in the provision of novel radiation shield devices which can be custom fabricated to fit the particular needs of a given application, in order to minimize installation difficulties while maximizing the effective dosage exposure reductions ultimately achieved. Other important but more specific objects of the invention reside in the provision of custom built, coated (e.g., painted) sheet lead shielding materials which: My radiation shields are also advantageously provided in specially designed and fabricated encapsulated (e.g, stainless steel) sheet lead materials which have additional important and more specific objectives, in that they: Coated radiation shields fabricated as described herein can be custom built, and specially designed and fabricated, and which are: Stainless steel encapsulated lead shields fabricated as described herein can be custom build, and specially design and fabricated, and which are: |
052723496 | abstract | A compact design of a shielded housing for transport and insertion of radioactive sources into wells for use in measuring properties of materials in vessels and the like is disclosed in this invention. The apparatus features a perforated belt which is wound on a drum. The sources are temporarily supported within the shielded housing and can be moved into place and mounted to the belt at the desired locations. Safety features are provided to avoid dropping the sources into the well and to temporarily support the belt so that the transport housing can be removed. The invention also provides for adjustment of the mounting height of the source or sources mounted to the belt. |
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abstract | A nuclear fuel composition includes a nuclear fissile material and a neutron-absorption material that adjoins the nuclear fissile material. The nuclear fuel composition may be used in a nuclear reactor, such as a thermal reactor. |
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055747582 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS Referring to FIGS. 1 and 2, a detecting device is shown, which comprises a primary detector 1 and a secondary detector 2 located in mutually opposed manner relative to the axis of a pipe P, e.g., a coolant pipe, through which for example, water of a non-reproductive cooler (not shown) flows, and a shield detector 3 surrounding substantially the whole perimeter of the primary detector 1 except for its portion facing the pipe. The primary detector 1 is constructed preferably of a semiconductor detector or a scintillation detector. The secondary detector 2 and the shield detector 3 are constructed each of one or more scintillation detectors. The semiconductor detector to be used includes, for example, HP Ge (high-purity germanium) detector, Si(Li) (lithium drift silicon) detector, CdTe (cadmium telluride) detector, GaAs (gallium arsenide) detector, HgI.sub.2 (mercuric iodide) detector, etc. The scintillation detector usable for this invention includes, for example, NaI(T1) (sodium iodide activated by tallium) detector, CsI(T1) (cesium iodide activated by thallium) detector, Bi.sub.4 Ge.sub.3 O.sub.12 (bismuth germanate known as BGO) detector, or the like. The secondary detector 2 is configured in a sector-form so as to surround a half to one-third of the periphery of the coolant pipe P. The primary, secondary, and shield detectors 1,2,3 are surrounded by lead shields 4 except for their directions in which gamma-rays and annihilation gamma-rays in the primary water are incoming and detected so that the incident gamma-rays and annihilation gamma-rays may be collimated and the incident dose of the gamma-rays and annihilation gamma-rays may be restricted. The detecting device including the primary detector 1, the secondary detector 2 and the shield detector 3 is connected with an anticoincidence circuit 11 for operating an anticoincidence counting between the primary and secondary detectors 1,2 and between the primary and shield detectors 1,3, and a multichannel pulse height analyzer 12, thus forming a gamma-ray spectrometric system as a whole, as shown in FIG. 3 in which essential components only are depicted with other components omitted since they are well-known per se in the art. In the gamma-ray spectrometry system thus constructed, the primary detector 1 serves to detect photons of gamma-rays from .sub.131 I, .sub.60 Co, and the like as intended as well as photons of the one annihilation gamma-rays as pulses; the secondary detector 2 serves to detect photons of the other annihilation gamma-rays as pulses; and the shield detector 3 serves to detect photons of the gamma-rays, which are Compton-scattered and escaped from the primary to the shield detectors, as pulses. When the photons of the gamma-rays or annihilation gamma-rays are detected on a germanium detector, an interaction of the photons with the germanium material yields gamma-ray spectra which usually include each a photoelectric peak due to full energy absorption event and a continuous sepctrum called Compton continuum or background due to once or twice scattering and subsequent escaping of the scattered photons outside the detector. Thus, the full energy absorption event is never attended with the escaping of scattered photons. The photopeak is utilized to determine the gamma-ray energy which is important for identification and quantitative determination of a radionuclide whereas the Compton backgound is a disturbing factor for the gamma-ray spectrometric measurement. Consequently, when the escaped photons are detected coincidently by means of the shield detector 3 and the primary detector 1 and an anticoincidence counting is operated, that tends to reject selectively the Compton scattering events only without affecting the full energy events. Further simultaneously when the annihilation gamma-rays are coincidently detected on the primary detector 1 and the secondary detector 2, and an anticoincidence counting is operated, that assists in rejecting selectively the events due to the annihilation gamma-rays without affecting the full energy events. More specifically, the rejection is conducted by passing the pulses from the primary detector 1 through electron gates of the anticoincidence circuit 11 which are adapted to be closed when pulses are detected on the secondary detector 2 and the shield detector 3, coincident with the detection on the primary detector 1. The rejection by anticoincidence counting operation yields the result that the annihilation gamma-rays and Compton background gamma-rays are significantly reduced, which enables it to increase the photopeak-to-background ratio in the spectrum and accordingly, to determine the count numbers of the intended gamma-rays with more precision. In the multichannel pulse height analyzer 12, the pulses detected by conversion of the radiation energy to voltage or current in proportion to the energy are divided into thousands of intervals (namely, multichannels) over the whole voltage or current pulse range, and number ratios of the pulses belonging to the respective channels are determined, thus yielding an energy distribution of the gamma-rays, i.e. gamma-spectra, from which count numbers of the intended gamma-rays are determined. One example of a method of the invention will be explained when applied to primary water of a nuclear reactor, i.e. non-reproductive cooler by fitting a coolant pipe P connecting to the cooler on its inflow side with the detecting device as described above including the primary, secondary, and shield detectors 1,2,3. As the primary detector 1, a germanium detector was used, which had a good energy resolution having a half band width of up to 2.0 KeV when 1.33 MeV gamma-ray of .sup.60 Co was taken as a standard and a counting efficiency of at least 75%. The shield detector used has such dimensions that make the Compton background (continuum) in the 131I area of the spectrum smaller than 1/10 of that without Compton suppression. When pulses from the primary detector 1 were counted in anticoincidence with pulses from the secondary detector 2 and the shield detector 3 by the operation of the anticoincidence circuit 11, the annihilation gamma-rays and Compton gamma-rays could be significantly diminished. Then, count numbers of the gamma-rays from .sup.131 I, .sup.60 Co, and others in the primary detector 1 were determined from the resulting gamma-spectra by analysis with the multi-channel pulse height analyzer 12. As a result, the detection limit of I gamma-ray area was enhanced to less than 1.5 Bq/cm.sup.3, more than 10 times as high as that (15 Bq/cm.sup.3) of a conventional method without reduction of the annihilation gamma-rays. This invention has been so far described, by way of example, with a primary water of a nuclear reactor, but the method can be naturally used for the analysis of: steam of a secondary system (from a steam generator), drain water of a primary coolant, chemical analysis of a primary coolant, etc. in the nuclear power field. However, this invention is also applicable to other fields, namely, researches in high energy physics, micro-analysis in accelerator engineering, etc. As described above, the conspicuous elevation of the detection limits of gamma-rays makes it possible to conduct continuous measurement of highly low-concentrations of radionuclides in primary water of a nuclear reactor, with the result that security of the nuclear reactor can be ensured by early detection of leakage of the fuel assembly. Further, it is possible to decrease the frequency of chemical analysis for detecting the leakage which has been hitherto performed in nuclear power plants. |
061223395 | claims | 1. In a reactor operated in accordance with operationally dependent input parameters, a method of operating the reactor which is unstable as a result of an oscillation of an internal physical variable, which comprises: measuring the physical variable by detecting a first oscillation peak of the physical variable, detecting the physical variable during a first oscillation period following the first oscillation peak, by subsequently detecting at least a second oscillation peak of the physical variable, and detecting the physical variable during at least a second oscillation period following the at least second oscillation peak and calculating at least one measured value for a rate of increase of the oscillation peaks detected during the first and the at least second oscillation periods by calculating a difference between the first oscillation peak of the physical variable and the at least second oscillation peak of the physical variable; and after the first and the at least second oscillation periods, deciding, in dependence on the measured value, whether a reactor operation is to be continued for a given number of oscillation periods with unchanged input parameters. 2. The method according to claim 1, which comprises defining a noise limit value and a threshold value for the oscillations higher than the noise limit value, and performing the calculating step after extreme values of the oscillation have exceeded the noise limit, and performing the deciding step only after the extreme values have exceeded the threshold value. 3. The method according to claim 1, which comprises defining a threshold value as a function of the measured value of the rate of increase of the oscillation, and intervening in an operational control of the reactor if extreme values of the oscillation exceed the threshold value. 4. The method according to claim 1, which comprises defining a threshold value for the rate of increase and initiating a stabilization strategy if the rate of increase exceeds the threshold value. 5. The method according to claim 1, which comprises defining a number of oscillations in dependence on the rate of increase, and intervening in a control of the reactor only if the oscillation of the physical variable persists for the number of oscillations. 6. The method according to claim 1, which comprises defining a plurality of stabilization strategies, each in dependence on a given rate of increase. 7. The method according to claim 6, wherein the defining step includes defining a low-ranking alarm stage stabilization strategy, which comprises intervening in the control of the reactor by preventing an increase in reactor power and blocking a removal of absorber elements from the reactor core. 8. The method according to claim 6, wherein the defining step includes defining stabilization strategies in variously ranked alarm stages including a first higher-ranking alarm stage, the first higher-ranking alarm stage intervention strategy comprising reducing a reactor power by slowly inserting absorber elements into the reactor core. 9. The method according to claim 6, wherein the defining step includes defining stabilization strategies in variously ranked alarm stages including a second higher-ranking alarm stage, the second higher-ranking alarm stage intervention strategy comprises rapidly inserting a plurality of control rods into the reactor core. 10. The method according to claim 6, wherein the defining step includes defining stabilization strategies in variously ranked alarm stages including a highest-ranking alarm stage, the highest-ranking alarm stage intervention strategy comprising reducing a power of the reactor to zero as rapidly as possible. 11. The method according to claim 1, which comprises combining a plurality of regions distributed over the reactor core into a system, and wherein the measuring and calculating steps comprise ascertaining a respective local measured value and a measured value for the rate of increase in each region of the system, and preliminarily deciding, with the measured values for each region, whether a stabilization strategy is to be initiated, and initiating the stabilization strategy only after a corresponding decision has been made in a minimum number of preliminary decisions. 12. The method according to claim 11, which comprises performing the preliminarily deciding step redundantly for a plurality of different systems made up of regions distributed over the reactor core, and blocking the stabilization strategy so long as a decision in favor of initiating the stabilization strategy has not been taken in a predefined minimum number of systems. 13. The method according to claim 12, which comprises assigning the regions to the systems in such a way that adjacent regions belong to mutually different systems, and each region belongs to only one system. 14. The method according to claim 11, which comprises providing a plurality of sensors in each region for ascertaining the measured values, and utilizing in each region only a predefined minimum number of the sensors in the deciding step. |
045541280 | claims | 1. Apparatus for determining the integrity and composition of TIG welds of an end plug on a sealed nuclear reactor fuel rod, comprising: (a) an ultrasonic weld inspection system for void detection with multiple transducers and having an immersion tank stuffing box with a seal for receiving said end plug on said fuel rod, said transducers disposed in said stuffing box to interrogate said welds of said end plug; (b) means for longitudinally moving said end plug into and out from said stuffing box; (c) an X-ray fluorescent spectrograph calibrated to detect tungsten and having a shielded counting chamber with an orifice for receiving said end plug on said fuel rod; (d) means for axially moving said end plug into and out from said counting chamber; (e) means for transporting said fuel rod between said stuffing box and said counting chamber; and (f) means for controlling, in a predetermined manner, said end plug axial moving means, said X-ray fluorescent spectrograph, said fuel rod transporting means, said end plug longitudinal moving means, and said ultrasonic weld inspection system. (a) a first ultrasonic weld inspection system for void detection with a first set of multiple transducers and having a first immersion tank stuffing box with a first gland seal for receiving said top end plug on said fuel rod, said first set of transducers disposed in said first stuffing box to interrogate said girth and seal welds of said top end plug; (b) means for longitudinally moving said top end plug into and out from said first stuffing box; (c) an X-ray fluorescent spectrograph calibrated to detect tungsten and having a shielded counting chamber with an orifice for receiving said top end plug on said fuel rod; (d) means for axially moving said top end plug into and out from said counting chamber; (e) a second ultrasonic weld inspection system for void detection with a second set of multiple transducers and having a second immersion tank stuffing box with a second gland seal for receiving said bottom end plug on said fuel rod, said second set of transducers disposed in said stuffing box to interrogate said girth weld of said bottom end plug; (f) means for lengthwise moving said bottom end plug into and out from said second stuffing box; (g) means for transporting said fuel rod between said first stuffing box, said counting chamber, and said second stuffing box; and (h) means for controlling, in a predetermined manner, said top end plug axial moving means, said X-ray fluorescent spectrograph, said fuel rod transporting means, said top end plug longitudinal moving means, said first ultrasonic weld inspection system, said bottom end plug lengthwise moving means, and said second ultrasonic weld inspection system. 2. The apparatus of claim 1, wherein said end plug longitudinal moving means includes a first group of driven rollers each having a circumferential groove disposed to receive said fuel rod, and said end plug axial moving means includes a second group of driven rollers each having a circumferential channel disposed to receive said fuel rod. 3. The apparatus of claim 2, wherein said two groups of rollers are parallel, and wherein said fuel rod transporting means includes a walking beam for carrying said fuel rod beween said grooves of said first group of rollers and said channels of said second group of rollers. 4. The apparatus of claim 1, wherein said fuel rod transporting means includes a walking beam. 5. The apparatus of claim 1, wherein said controlling means also includes means for generating a first reject signal when said X-ray fluorescent spectrograph detects tungsten in an amount exceeding a precalculated value. 6. The apparatus of claim 1, wherein said controlling means also includes means for generating a second reject signal when said ultrasonic weld inspection system detects a weld void exceeding a preselected magnitude. 7. The apparatus of claim 1, wherein said controlling means includes a microprocessor. 8. Apparatus for inspecting TIG girth and seal welds of a top end plug and a girth TIG weld of a bottom end plug, both end plugs on a sealed nuclear reactor fuel rod, said apparatus comprising: |
summary | ||
047956054 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS In FIGS. 2 and 3, a central support stay 1 according to a first embodiment of the present invention comprises a central support portion 1f extending along the axis of the central support stay 1 and a plurality of radial members 1b connected to the central support stay 1 and radially extending from the central support portion 1f and circumferentially spaced from each other at predetermined angles. The central support portion 1f and the radial members 1b are integrally formed. The number of radial members 1b is preferably equal to the number of toroidal coils 2 circumferentially spaced from each other as shown in FIG. 1. The radial outer end surface of each radial member 1b constitutes a load portion 1a to which a centripetal force F from each toroidal coil 2 is applied. A plurality of poloidal coils 3 are circumferentially disposed around the outer surfaces of the radial members 1b of the central support stay 1 and are spaced from each other along the axis of the central support stay 1. In the first embodiment mentioned above, the centripetal force F generated in each toroidal coil 2 is applied to the load portion 1a of each radial member 1b. Each of the radial members 1b radially extend outwards from the central support portion 1f and have enough radial rigidity to withstand the centripetal force of each toroidal coil 2. Since the radial members 1b are spaced from each other at predetermined angles in the circumferential direction of the central support stay 1, the electrical resistance between the radial members 1b in the circumferential direction is higher than that in the conventional apparatus, and it is unnecessary to dispose electrically insulating materials between the circumferentially segmented portions as in the conventional apparatus. Further, the structure of the central support stay 1 mentioned above is simple. As shown in FIG. 4, lead wires 3a for supplying an electric current to each of the poloidal coils are attached to the poloidal coils themselves which are then easily fitted onto the outer circumference of the central support stay 1 along the axial direction thereof by locating the lead wires 3a in the respective clearances 1c between the radial members 1b so that there is no need to perform operations for disposing holes and brazing as in the conventional apparatus. The first embodiment mentioned above refers to a case in which the number of radial members 1b is equal to the number of toroidal coils 2. However, the number of radial members 1b may be one half the number of toroidal coils 2. In this case, as shown in FIG. 5, a force transmitting member 6 may be preferably disposed at the radial outer end of each radial member 1b such that the centripetal forces from two toroidal coils 2 are transmitted to one radial member 1b. Similar effects can be obtained even when the number of radial members 1b is a whole fraction of the toroidal coils. In FIG. 4, although the lead wires 3a for supplying an electric current to the poloidal coils 3 are disposed therein, the assembly of the apparatus is similarly facilitated even when an opening end of a conduit for cooling, e.g., a superconduction coil used as each of the poloidal coils is disposed in each poloidal coil. In the first embodiment, although the central support stay is integrally formed, as shown in FIG. 6, a central support stay 50 may be constituted of a plurality of circumferentially laminated layers 52 between which electrically insulating materials are disposed, thereby further increasing the electrical resistance in the circumferential direction of the central support stay. Furthermore, as shown in FIG. 7, a central support stay 60 may comprise a central support portion 62 and a plurality of radial members 64 radially extending from the central support portion 62 and each constituted of a plurality of layers laminated in the radial direction of the central support stay, thereby increasing the electrical resistance with respect to eddy currents radially generated in the radial members 64. In FIG. 7, the central support portion 62 may be constituted by a plurality of layers laminated in the axial direction of the central support stay. The embodiments shown in FIGS. 6 and 7 have an advantage in that it is unnecessary to use a large-sized material as the central support stay. The central support stay may be made of metal or a highly electrically resistant material such as ceramic, FRP, etc. In the case of such highly resistant material, the central support stay can be integrally formed and no eddy current is generated in the central support stay. As mentioned above, in a nuclear fusion apparatus according to the first embodiment, a central support stay can sufficiently withstand the centripetal forces of toroidal coils disposed around the central support stay and has a high electrical resistance in the circumferential direction of the central support stay, and has a simple structure in which it is easy to dispose lead wires for supplying electric current to each of the poloidal coils. FIGS. 8 and 9 show a fifth embodiment of the present invention. A central support stay 11 comprises central support portions 12 disposed at the top and bottom ends of the stay 11, and a plurality of radial members 13 radially disposed between the upper and lower support portions 12 and radially movable within respective radially arranged holders 70 with respect to the support portions 12 as described later. The radial members 13 are circumferentially spaced from each other at predetermined angles and each of them may be constituted by a member similar to a piston. The sliding surfaces 13a of the radial members 13 and the contact surfaces 12a of the support portions 12 contacting the radial members 13 seal in oil tight cooperation with each other. An oil 15 is supplied into a central space defined by the support portions 12, the holder 70 and the radial members 13 and extending along the axis of the central support stay 11. The radial members 13 can be respectively radially moved by an oil pressure device for pressurizing the oil 15 within the central space. A plurality of grooves 14 for receiving poloidal coils 3 are circumferentially disposed at the radial outer ends of the radial members 13 and are spaced from each other along the axis of the central support stay 11. The poloidal coils 3 are held within the respective grooves 14 of the radial members 13 and can be attached to the respective radial members 13. The radial outer end surface 13b of each radial member 13 is adjacent each toroidal coil 2 and can withstand the centripetal force of each toroidal coil 2 as described later. Clearances 17 are formed between the radial members 13 in the circumferential direction of the central support stay 11 and are used for receiving pipes for the poloidal coils 3 through the clearances 17. The oil 15 in the central space can be heated by a heater 18 disposed in each radial member 13. In the fifth embodiment constructed as above, each of the radial members 13 is radially movable with respect to the support portions 12 by the oil pressure device. Accordingly, the radial members 13 are first moved inwards such that the poloidal coils 3 are circumferentially disposed around the grooves 14 of the radial members 13 in the central support stay 11. The radial members 13 are then moved outwards by the oil pressure device for pressurizing the oil 15 so that the poloidal coils 3 can be respectively fitted into the grooves 14 of the radial members 13. The poloidal coils 3 within the grooves 14 are thus respectively supported by the radial members 13. When the pressure applied to the oil 15 is increased by the oil pressure device, the outer end surfaces 13b of the radial members 13 respectively contact and press the toroidal coils 2 outwards so that the radial members 13 can withstand the centripetal forces generated in the toroidal coils 2 at the outer end surfaces 13b thereof. In the fifth embodiment mentioned above, when the nuclear fusion apparatus is operated, after a predetermined pressure has been applied to the oil 15 by the oil pressure device such that the radial members 13 press the toroidal coils 2 outwards to withstand the centripetal forces thereof, the clearances 17 are filled with, e.g., liquid nitrogen to solidify the oil 15 and the operation of the oil pressure device is then stopped. Thus, the poloidal coils 3 can be more efficiently cooled and thereby the vaporisation of liquid helium filling the grooves 14 therewith for cooling the poloidal coils 3 can be lowered. In the maintenance of the nuclear fusion apparatus, the liquid nitrogen is removed from the clearances 17 and the solidified oil 15 is reliquefied by heating the oil by the action of the heater 18 disposed in each radial member 13. The pressure of the oil 15 is then decreased by the oil pressure device so that the radial members 13 can be moved inwards, thereby releasing the poloidal coils 3 from out of the grooves 14 of the radial members 13. Both the central support stay 11 and the poloidal coils 3 can be integrally taken out with slight clearances between the radial members 13 and the toroidal coils 2, instead of moving the radial members 13 completely into the portion between the upper and lower support portions 12. In the fifth embodiment, the oil 15 is solidified to hold the radial members 13 in predetermined positions, but other means for holding the radial members 13 in positions may be used. For example, as in a sixth embodiment shown in FIG. 10, a bar member 19 inserted into holes formed in each radial member 13 and each support portion 12 may be used to fix each radial member 13 to each support portion 12. Furthermore, instead of solidifying the oil 15, the operation of the heater 18 may be adjusted so as to maintain the oil 15 in a liquid phase at any time so that a pressure can be applied to the oil 15 even when the apparatus is operated. In this case, as in a seventh embodiment shown in FIG. 11, a sensor 20 for detecting the pressure of the oil 15 attached to the oil pressure device may be disposed in the central support stay 11 to measure the pressure of the oil in the central support stay 11 during operation of the apparatus such that the oil pressure can be suitably maintained. Furthermore, an annunciator or an emergency stopping device may be disposed as a safe system for any abnormal operation of the apparatus. Although the radial members 13 are moved radially by the oil pressure, any other known means for withstanding the centripetal forces of the toroidal coils 2 may be used to move the radial members radially outwards. In the fifth to seventh embodiments, each radial member in the central support stay is radially movable and has a groove at the radial outer end thereof for receiving a poloidal coil and they can be pressed outwards with a predetermined pressure for withstanding the centripetal force generated by the toroidal coils disposed around the central support stay. By such a construction, the electrical insulation in the circumferential direction of the central support stay is not reduced and the poloidal coils can be disposed without any additional means for positioning the poloidal coils around the central support stay. Furthermore, the assembly of the apparatus is simplified and the strength of the central support stay can be increased. FIGS. 12 to 14 show an eighth embodiment of the present invention. A central support stay 110 comprises a plurality of support stay members 111 segmented in the axial direction of the central support stay 110. Each of the support stay members 111 has a central support segment 140 and a plurality of radial segments 111b radially extending from the central support segment 140 and circumferentially spaced from each other at predetermined angles, thereby forming clearances 111c respectively between the radial segments 111b in the circumferential direction. The radial outer ends 111a of each radial segment 111b constitute portions for withstanding the centripetal force from each toroidal coil 2. A circular projected portion 111e and a circular recessed portion 111f are respectively disposed in the top and bottom of each support stay member 111. The projected portion 111e of each lower support stay member 111 is fitted into the recessed portion 111f of each upper adjacent support stay member 111 along the axis of each support stay member 111 to integrate the support stay members with each other in the assembly of the apparatus. The upper contact surface 111d of each support stay member 111 contacts the bottom surface of each upper adjacent support stay member 111 when the support stay members are assembled. A step 111h is disposed in the upper radial outer end portion of each radial segment 111b and has a vertical surface portion 111m extending in the axial direction of each support stay member 111 and a horizontal or flat surface portion 111g extending in the radial direction. Each of the poloidal coils 3 are arranged on the horizontal surface portion 111g of each step 111h as described later. Each of the support stay members 111 have a through hole 111j axially extending therethrough in the center of each support stay member 111 for receiving a fastener such as a bolt 112. When the support stay members 111 are assembled with each other in the axial direction, the through holes 111j of the support stay members are set to be aligned with each other along the axis of the central support stay 110. A nut 113 is screwed onto the bolt 112 to axially fasten the assembled support stay members 111 to each other. The uppermost support stay member 114 does not have the circular projected portion 111e nor the step 111h in the upper portion thereof. The lowermost support stay member 115 does not have the circular recessed portion 111f in the bottom thereof. The axial lengths of the uppermost and lowermost support stay members 114 and 115 are shorter than those of the support stay members disposed therebetween. The number of radial segments 111b is preferably equal to the number of toroidal coils 2 as in the first embodiment. When the poloidal coils 3 are disposed around the central support stay 110, the poloidal coils 3 are respectively located on the horizontal surfaces 111g of the steps 111h of the radial segments 111b in the lower support stay members 111 and the upper support stay members 111 are next assembled on the lower support stay members 111 such that the projected portion 111e of each lower support stay member 111 is fitted into the recessed portion 111f of each upper adjacent support stay member 111. The bolt 112 is then inserted into the central through hole 111j of each of the assembled support stay members 111 and fastens the assembled support stay members 111 to each other in cooperation with the nut 113 screwed onto the bolt 112. The poloidal coils 3 are firmly fixed between the assembled support stay members. As shown in FIG. 12, centripetal forces F generated in the toroidal coils 2 are transmitted to the radial segments 111b of each support stay member 111 through the radial outer ends 111a thereof contacting the toroidal coils 2. The radial segments 111b radially extending outwards each have high rigidity in the radial direction of the support stay member 111 so that the radial segments 111b can withstand the centripetal forces of the toroidal coils 2 as in the first embodiment. In the eighth embodiment shown in FIGS. 12 to 14, effects similar to those obtained in the first embodiment can be obtained. In the eighth embodiment, the through hole 111j for receiving the bolt 112 is disposed in the central portion of each support stay member 111 to fasten the support stay members 111 to each other in the axial direction thereof. However, as in a ninth embodiment shown in FIGS. 15 and 16, a through hole 122 may be axially disposed in each radial segment 111b' of each support stay member 111' such that the axis of each through hole 122 is located on a common circle 121 the center of which is aligned with the axis of the central support stay 110'. Each of the through holes 122 receive a fastener such as a bolt for fastening the support stay members 111' to each other in combination with a nut. In this case, in the assembly of the support stay members, the support stay members 111' can be easily positioned in the circumferential directions thereof and fastened to each other in the axial directions thereof in a fashion inherently stronger than that of the eighth embodiment using only one bolt. In addition, in the eighth embodiment, means for preventing the support stay members from being circumferentially rotated with respect to each other may be disposed between the adjacent support stay members to position the support stay members in the circumferential directions thereof. According to the eighth and ninth embodiments, a central support stay comprises a plurality of support stay members segmented in the axial direction of the stay, and each of the support stay members has a central support segment and a plurality of radial segments radially extending therefrom and circumferentially spaced from each other at predetermined angles and provided with steps for receiving poloidal coils. The respective support stay members are fastened and fixed to each other by a fastener inserted into through holes extending through the respective support stay members along the axial direction thereof. Therefore, the poloidal coils are firmly fixed to the support stay members therebetween and the support stay members can sufficiently withstand the centripetal forces of the toroidal coils. |
claims | 1. A method of irradiating a target specimen within a nuclear reactor for at least one fuel cycle, to produce at least one commercial radioisotope, the method comprising steps of:enclosing the target specimen within an elongated tubular housing having an axis along its elongated dimension,the target specimen being nuclear reactor transmutable to produce the at least one commercial radioisotope,the elongated tubular housing being closed at a forward end and capped at a rearward end to form a target specimen chamber therebetween within an interior of the elongated tubular housing, andthe elongated tubular housing being sized to slide within an instrument thimble of a nuclear fuel assembly, with the rearward end structured to be driven by a drive cable of an existing moveable in-core detector system;positioning the target specimen at a preselected axial position within the elongated tubular housing, wherein the target specimen is captured between a forward axial position plug and a rear axial position plug, wherein the forward and rear axial position plugs are structured to seat against an interior wall of the elongated tubular housing to hold the target specimen at the preselected axial position within the elongated tubular housing;attaching the rearward end to the drive cable;driving the target specimen positioned within the elongated tubular housing into an instrument thimble of a selected nuclear fuel assembly within a core of a nuclear reactor;leaving the target specimen within the instrument thimble for the remainder of a fuel cycle of the core, wherein the target specimen while in the instrument thimble is nuclear reactor transmuted to produce the at least one commercial radioisotope;withdrawing the elongated tubular housing with the at least one commercial radioisotope therein from the core at the end of the fuel cycle;removing the selected fuel assembly from the core;while the selected fuel assembly is removed from the core, reinserting the elongated tubular housing with the at least one commercial radioisotope therein at least partially into the core; andwhile the elongated tubular housing is at least partially in the core, dislodging from the drive cable at least a portion of the elongated tubular housing that has the at least one commercial radioisotope therein. 2. The method of claim 1, wherein the dislodging step cuts the elongated tubular housing around a circumference. 3. The method of claim 2, including a step of transferring the at least a portion of the elongated tubular housing that has the at least one commercial radioisotope therein under water to a spent fuel pool. 4. The method of claim 3, comprising transferring the at least a portion of the elongated tubular housing in a building housing the spent fuel pool to a shielded package for shipment. 5. The method of claim 1, wherein positioning the isotope target specimen at a preselected axial position within the elongated tubular housing comprises positioning the target specimen between the forward axial position plug and the rear axial position plug both of which extend across the interior of the elongated tubular housing. 6. The method of claim 1, wherein the at least one commercial radioisotope comprises one or more materials selected from the group consisting of: Co-60, W-188, Ni-63, Bi-213, and Ac-225. 7. The method of claim 1, wherein the elongated tubular housing is constructed from zirconium or a zirconium alloy. 8. The method of claim 1, wherein the forward and rear axial position plugs maintain their axial position due to friction between interfacing surfaces on the axial position forward and rear plugs and the interior wall of the elongated tubular housing. 9. The method of claim 1, wherein the forward and rear axial position plugs maintain their axial position by fitting in slight recesses in the interior wall of the elongated tubular housing. 10. The method of claim 9, wherein the forward and rear axial position plugs have an upper and lower surface that extends substantially orthogonal to the axis with an outer, substantially circular wall extending between the upper and lower surface, wherein the axial dimension of the outer, substantially circular wall is sized to fit in one of the recesses. |
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047987006 | abstract | A gas-cooled high temperature reactor is provided having a core filled with spherical fuel elements, in combination with a graphite side reflector including at least one nose-like projection protruding radially into the reactor core from said graphite said reflector, the at least one nose-like projection including at least one vertically disposed cavity adapted to receive discrete absorber material elements introduced into said reactor core as well as a vertically disposed continuous opening which permits communication between said cavity and the core of the reactor, said opening having a maximum width adjacent said cavity which is less than the minimum dimension of said discrete absorber material elements in order to prevent passage of said elements into said continuous opening from said cavity. |
claims | 1. A system for expanding the diameter of a tubular disposed within a wellbore, comprising:an expandable tubular having an interior surface, andan expansion tool configured to fit within a perimeter defined by the interior surface, the expansion tool having a selectively expandable portion, wherein the selectively expandable portion imparts a radial expansion force against the interior surface to drive the expandable tubular to an expanded state, wherein the selectively expandable portion comprises a plurality of pistons, wherein the pistons actuate under the influence of a biasing member, and wherein the pistons comprise subsystem members positioned to rotatably engage the biasing member. 2. A system for expanding the diameter of a tubular disposed within a wellbore, comprising:an expandable tubular having an interior surface, andan expansion tool configured to fit within a perimeter defined by the interior surface, the expansion tool having a selectively expandable portion, wherein the selectively expandable portion imparts a radial expansion force against the interior surface to drive the expandable tubular to an expanded state, wherein the selectively expandable portion comprises a plurality of pistons, wherein the pistons actuate under the influence of a biasing member, and wherein the biasing member travels upwardly through the wellbore. 3. The system as recited in claim 2, further comprising a wireline adapted to engage the biasing member, the wireline being insertable into the wellbore under influence of a fluid. 4. The system as recited in claim 3, wherein the wireline comprises a plurality of flanges adapted to receive the fluid. 5. A system for expanding the diameter of a tubular disposed within a wellbore, comprising:an expandable tubular having an interior surface; andan expansion tool configured to fit within a perimeter defined by the interior surface, the expansion tool having a selectively expandable portion, wherein the selectively expandable portion imparts a radial expansion force against the interior surface to drive the expandable tubular to an expanded state, wherein the expansion tool comprises an inflatable member disposed along a central mandrel. 6. The system as recited in claim 5, wherein the inflatable member comprises a plurality of inflatable members and inflates via a liquid. 7. A system for expanding the diameter of a tubular disposed within a wellbore, comprising:an expandable tubular having an interior surface; andan expansion tool configured to fit within a perimeter defined by the interior surface, the expansion tool having a selectively expandable portion, wherein the selectively expandable portion imparts a radial expansion force against the interior surface to drive the expandable tubular to an expanded state, wherein the expansion tool comprises a compressible elastomer. 8. A system for expanding the diameter of a tubular disposed within a wellbore, comprising:an expandable tubular having an interior surface; andan expansion tool configured to fit within a perimeter defined by the interior surface, the expansion tool having a selectively expandable portion, wherein the selectively expandable portion imparts a radial expansion force against the interior surface to drive the expandable tubular to an expanded state, wherein the expansion tool comprises a compressible spring, the spring being adapted to radially expand during transition from a compressed configuration to an expended configuration. 9. A system for expanding the diameter of a tubular disposed within a wellbore, comprising:an expandable tubular having an interior surface; andan expansion tool configured to fit within a perimeter defined by the interior surface, the expansion tool having a selectively expandable portion, wherein the selectively expandable portion imparts a radial expansion force against the interior surface to drive the expandable tubular to an expanded state, the expansion tool further comprising a roller, wherein the roller comprises elliptical members having an interior engagement surface; andfurther comprising an axle, wherein the interior engagement surface of the roller travels along a circumference of the axle. 10. A system for expanding the diameter of a tubular disposed within a wellbore, comprising:an expandable tubular having an interior surface; andan expansion tool configured to fit within a perimeter defined by the interior surface, the expansion tool having a selectively expandable portion, wherein the selectively expandable portion imparts a radial expansion force against the interior surface to drive the expandable tubular to an expanded state, wherein the expansion portion comprises a plurality of expandable discs. 11. The system as recited in claim 10, further comprising a removable sleeve disposed about the expandable discs, wherein the sleeve retains the expandable discs in a compressed configuration. 12. A system for expanding the diameter of a tubular disposed within a wellbore, comprising:an expandable tubular having an interior surface; andan expansion tool configured to fit within a perimeter defined by the interior surface, the expansion tool having a selectively expandable portion, wherein the selectively expandable portion imparts a radial expansion force against the interior surface to drive the expandable tubular to an expanded state, wherein the expansion tool comprises a first rotating member coupled to a second rotating member, wherein rotation of the first member about the second member provides the radial expansion force. 13. A system for expanding the diameter of a tubular disposed within a wellbore, comprising:an expandable tubular having an interior surface; andan expansion tool configured to fit within a perimeter defined by the interior surface, the expansion tool having a selectively expandable portion, wherein the selectively expandable portion imparts a radial expansion force against the interior surface to drive the expandable tubular to an expanded state, wherein the expansion tool comprises a plurality of block members, wherein at least one of the plurality of block members is adapted to travel radially outward in response to an axial compressive force. 14. An expansion system to expand a tubular disposed in a wellbore, comprising:an expansion mechanism sized for deployment within the interior of the tubular, the expansion mechanism comprising a radially expandable portion, the radially expandable portion being configured to enable selective expansion of the tubular to an expanded state by imparting a force directed radially against the tubular, wherein the expansion mechanism comprises an inflatable member disposed along a supporting mandrel. 15. An expansion system to expand a tubular disposed in a wellbore, comprising:an expansion mechanism sized for deployment within the interior of the tubular, the expansion mechanism comprising a radially expandable portion, the radially expandable portion being configured to enable selective expansion of the tubular to an expanded state by imparting a force directed radially against the tubular, wherein the expansion mechanism comprises an expansion plate biased in a radially outward direction with respect to an axis of the wellbore. 16. An expansion device for expanding a tubular within a wellbore, comprising a mandrel having a stepped profile oriented to engage an interior surface of the tubular, the stepped profile being formed of adjacent stages, each stage having a smaller diameter than the preceding stage along the direction of movement of the mandrel during expansion. 17. The expansion device as recited in claim 16, wherein the stepped profile extends along a portion of the mandrel in an axial direction. 18. A method for expanding a tubular having contracted and expanded states, comprising:disposing a tubular in a contracted state within a wellbore;disposing an expansion tool at least partially within an interior region of the contracted tubular; andactivating an expansion portion of the expansion tool such that the expansion portion imparts a radial force on the tubular sufficient to transition the tubular to a radially expanded configuration, wherein activating comprises inflating a plurality of tubes. 19. A method for expanding a tubular having contracted and expanded states, comprising:disposing a tubular in a contracted state within a wellbore;disposing an expansion tool at least partially within an interior region of the contracted tubular; andactivating an expansion portion of the expansion tool such that the expansion portion imparts a radial force on the tubular sufficient to transition the tubular to a radially expanded configuration, wherein activating comprises rotating the expansion member. 20. A method for expanding a tubular having contracted and expanded states, comprising:disposing a tubular in a contracted state within a wellbore;disposing an expansion tool at least partially within an interior region of the contracted tubular; andactivating an expansion portion of the expansion tool such that the expansion portion imparts a radial force on the tubular sufficient to transition the tubular to a radially expanded configuration, wherein activating comprises removing a sleeve positioned to restrict expansion of the expansion portion. 21. A method for expanding a tubular having contracted and expanded states, comprising:disposing a tubular in a contracted state within a wellbore;disposing an expansion tool at least partially within an interior region of the contracted tubular; andactivating an expansion portion of the expansion tool such that the expansion portion imparts a radial force on the tubular sufficient to transition the tubular to a radially expanded configuration, wherein activating comprises compressing the expansion tool via an axial compressive force. |
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050733354 | abstract | A recirculation system is disclosed for driving reactor coolant water in an annular downcomer defined between a reactor vessel and a core shroud spaced radially inwardly therefrom. The system supplies feedwater to the vessel and to a turbopump disposed inside the downcomer. The turbopump in accordance with one embodiment of the present invention includes a stationary axle and a pump impeller rotatably joined thereto and having an inlet end for receiving the coolant water from the downcomer. An annular plenum surrounds the impeller for channeling feedwater to a plurality of turbine blades joined to the impeller for rotating the impeller for driving the coolant water. The impeller is lubricated solely by the feedwater upon rotation of the impeller about the axle. |
abstract | Disclosed is a beam monitor system in which signals outputted from a plurality of wires are divided in a multi-wire type monitor for measuring a beam profile of a charged particle beam, an identical number of the wires are grouped, the signals of the respective groups are taken out one piece by one piece to be connected with each other, and the number of the pieces, corresponding to a number of the wires belonging to the one group, are put together to be connected to a signal processor storing connection information. |
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abstract | A grid comprises an alternate assembly of radiation-permeable members and radiation-impermeable members which extend substantially parallel to the chest wall of a subject. When a radiation emitted from a radiation source is applied through a breast of the subject and the grid to a radiation detector, a radiation image of the breast is captured. While the radiation is being applied to the breast, the grid reciprocates in directions perpendicular to the direction in which the radiation-impermeable members extend. |
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abstract | The invention relates to a collimator (1) for limiting a bundle of high-energy rays (2), which is emitted by a substantially point-like radiation source (3) and directed towards a treatment object (20) and used in particular for the stereotactic conformation radiotherapy of tumors. According to the invention the collimator (1) comprises a plurality of diaphragm leaves (4, 4xe2x80x2) which are arranged opposite each other and which are made of a radiation-absorbing material and which, by means of drive mechanisms, can be moved into the optical path in such a way that the contours and/or exposure period of said optical path can be freely defined, the front edges (5, 5xe2x80x2) of the diaphragm leaves (4, 4xe2x80x2) being parallel to the optical path at all times. Avoiding penumbral shadows with this kind of collimator (1) is made considerably easier if the diaphragm leaves (4, 4xe2x80x2) consists of a rear partial element (6, 6xe2x80x2) which can be linearly displaced and a front partial element (7, 7xe2x80x2) which is hinged to same. Drive means adjust the front partial element (7, 7xe2x80x2) in accordance with the prevailing position of the rear partial element (6, 6xe2x80x2) in such a way that the front edges (5, 5xe2x80x2) are parallel to the optical path at all times. |
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abstract | A process for producing a daughter radionuclide from a parent radionuclide includes a) loading the parent radionuclide on a first solid medium contained in a generator and onto which the parent radionuclide is retained and whereby the daughter radionuclide is formed by radioactive decay of the parent radionuclide; b) eluting this medium with a A0 solution so as to recover a A1 solution comprising the daughter radionuclide; c) optionally adjusting the pH of the A1 solution so as to obtain a A1′ solution, d) loading this A1 or A1′ solution onto the head of a second solid medium contained in a chromatography column; e) first washing said second solid medium with a A2 solution; f) second washing said second solid medium with a A2′ solution; g) eluting the daughter radionuclide with a A3 solution. The first washing step is conducted from head to tail of the column and the second washing step and the second eluting step are conducted from tail to head of the column. |
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