Abstract: A tool for use in servicing an LPRM assembly of a nuclear reactor vessel includes a structural member having a first bore that is structured to receive an LPRM device associated with the LPRM assembly, a headpiece provided at a first end of the structural member, and a nut engaging assembly slideably mounted on the structural member. The headpiece has a plurality of projections structured to mate with a plurality of bores provided in a seal of the LPRM assembly to enable the seal to be removed, and the nut engaging assembly has a housing that defines a second bore and that has first and second nut engaging portions. The nut engaging assembly is free to slide along the structural member and over the headpiece to a position wherein the engaging portions extend beyond the headpiece so that they may be used to remove the assembly nut.

Abstract: A method and system for entropy coding can comprise, in response to detecting a first symbol combination comprising first run information indicating a first number of contiguous zero coefficients is greater than a cut-off-run value, assigning a first codeword to a first symbol combination, wherein the first codeword comprises an escape code from a first-level VLC table; and in response to a second symbol combination comprising second run information indicating a second number of contiguous zero coefficients is less than or equal to the cut-off-run value, assigning a second codeword to the second symbol combination, wherein the second codeword is from the first-level VLC table. The system and method can further comprise collecting coding statistics for a set of candidate symbol combinations and adjusting a mapping between codewords of the first-level VLC table and a subset of the set of candidate symbol combinations based on the coding statistics.

Abstract: A nuclear fission reactor, flow control assembly, methods therefor and a flow control assembly system. The flow control assembly is coupled to a nuclear fission module capable of producing a traveling burn wave at a location relative to the nuclear fission module. The flow control assembly controls flow of a fluid in response to the location relative to the nuclear fission module. The flow control assembly comprises a flow regulator subassembly configured to be operated according to an operating parameter associated with the nuclear fission module. In addition, the flow regulator subassembly is reconfigurable according to a predetermined input to the flow regulator subassembly. Moreover, the flow control assembly comprises a carriage subassembly coupled to the flow regulator subassembly for adjusting the flow regulator subassembly to vary fluid flow into the nuclear fission module.

Abstract: In a nuclear reactor in which a primary coolant is contained, the primary coolant moves upwardly from the core by an operation thereof. An annular steam generator is arranged in an upper side of the core into which the upwardly moving primary coolant flows and transfers heat in the primary coolant into water therein to generate a steam. A passage structure defines a coolant passage for the primary coolant to an outside of the core. The heat-transferred primary coolant in the annular steam generator flows downwardly in the coolant passage so as to flow into the core, thereby moving upwardly. A reactor vessel is arranged to surround the coolant passage so as to contain the core, the annular steam generator and the passage means therein.

Abstract: With this method it is possible to evaluate opposing friction forces occurring after a certain operating time on a mobile object in a guide. The method consists of measuring the variations in the speed of movement of the mobile object, of calculating the theoretical speed of movement without friction, of combining these two results and thereby deducing outside opposing forces through calculation of the acceleration of the mobile object. Particular application to the evaluation of additional friction forces acting on a mobile control cluster assembly in the core of a pressurized water nuclear reactor.

Abstract: A method for detecting the dropping of at least one control element in a reactor core of a power station by delaying signals output by detectors disposed along a fall path of the control element in such a way that they are approximately simultaneous. This improves a registration of the dropping of the control element by improving a ratio between a useful signal and a noise signal.

Abstract: A CEDM data acquisition system (40) receives analog coil-current signals from control element drive mechanism coils (18, 20); conditions the analog signals to remove noise induced therein by the electronic circuitry (44) associated with the control element drive mechanisms; digitizes the conditioned analog signals; displays the acquired test data; and records the data for future use. The inventive data acquisition system preferably has the ability to simultaneously measure, display and record coil-current data for all of the five coils associated with each CEDM rod-group with respect to time. Further, the inventive data acquisition system preferably allows the user to monitor, display and record data for up to eight CEDMs simultaneously. It also allows the measurement of rod-drop times during rod-drop testing of the CEDMs.

Abstract: A reactor protection system having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Each division performs independently of the others (asynchronous operation).

Abstract: A system for simultaneously testing at least two any control rod clusters contained within a reactor vessel, the system comprising a control rod drive mechanism attached to the control rod clusters for retracting the control rod clusters within said reactor vessel to a position suitable for testing. Electrical power means connected to the control rod drive mechanism for supplying electrical power to the control rod drive mechanism and for terminating the power to the control rod drive mechanism and, when terminated, causing said all said control rod clusters to fall into the reactor vessel. A rod position indicator attached to the control rod drive mechanism for monitoring the position of the control rod clusters; and computing means operatively connected to the rod position indicator and receiving a signal representing the fall time of each control rod cluster for generating an elapsed time profile of all the control rod clusters.

Abstract: A method for magnetic coupling compensating a rod position indication system having at least two rod position indicators each having induced noise comprising the steps of applying a sinusoidal current to a primary of the first rod position indicator for inducing a voltage on a first secondary of the first rod position indicator; receiving both the first secondary voltage from the first rod position indicator and a second secondary voltage, induced from the noise, of the second rod position indicator; and deriving a difference of the first and second secondary voltage for obtaining the magnetic coupling compensation.

Abstract: Certain of the nuclear fuel assemblies contained in a fast neutron nuclear reactor core have an integrated passive safety device.More specifically, said device is constituted by an annular block (48) forming at least the lower part of the upper neutron protection (42) of the assembly. The block (48) is formed by elements (50) of a neutron-absorbing material, embedded in a matrix (52) meltable at a temperature above the temperature of the sodium during the normal operation of the reactor. A perforated, inner envelope (54) mechanically secures the block (48), whilst enabling the elements (50) to drop in the assembly following the melting of the matrix (52).

Abstract: Segmented instrumentation tube including a locking sleeve for interlocking the segments of the instrumentation tube, so that the threaded ends of the instrumentation tube do not unthread when subjected to vibration, such an instrumentation tube being suitable for use in a nuclear reactor pressure vessel. The instrumentation tube has a first member having a threaded end portion that has a plurality of first holes circumferentially around the outside surface thereof. The instrumentation tube also has a second member having a threaded end portion that has a plurality of second holes circumferentially around the outside surface thereof. The threads of the second member are caused to threadably engage the threads of the first member for defining a threaded joint therebetween. A sleeve having an inside surface surrounds the end portion of the first member and the end portion of the second member and thus surrounds the threaded joint.

Abstract: A system for monitoring an operating condition of a control rod drive mechanism (CRDM) having a pressure housing erectly mounted on a reactor vessel head closure, a vertically movable control rod drive shaft accommodated in the housing and an electromagnetically motive latch mechanism includes an accelerometer fitted to a housing upper end and signal recording means for receiving output signals from the accelerometer through signal processing means and recording the processed output signals.

Abstract: An apparatus and method for scanning the outer profile of control rods in a control rod assembly includes a plurality of rotatably mounted ultrasonic transducers which are mounted for rotation about the axis of at least some of the control rods in the assembly. The assembly is moved in a linear path parallel to the axis of the control rods, past the rotating transducers for scanning the surfaces of the control rods in helical paths. By distributing the transducers in a selected configuration, all control rods in the assembly can eventually be scanned.

Abstract: The rod position indicator system of a pressurized nuclear reactor can be calibrated for one or more control rods without shutdown of the reactor. The procedure involves lowering reactor power to a level at which the control rod which appears to be misaligned can be fully inserted and its rod position indicator calibrated.

Abstract: A control rod drop monitoring method and apparatus for use in a nuclear reactor including dedicated timers and displays to measure and display the drop time for each control rod when a reactor system trip occurs. The timers and displays may be made from CMOS technology, and may have millisecond resolution, if desired. The timers are initiated upon receipt of one or more reactor trip signals and are halted at the closure of switches for detecting the fully inserted or other preselected position of the control rods in the reactor core. The monitor system is designed for continuous operation. Reactor trip signals and limit switch signals are fed to latch circuits following a trip event or limit switch activation and are unaffected by subsequent changes in the state of the limit switches or trip signals. A manual reset is required to clear rod drop times. Also, a test mode is provided wherein all of the timers can be started and stopped, thereby providing a functional test of 100% of the monitor circuitry.

Abstract: A nuclear fuel assembly has mechanical elements which constrain the flux thimble tube within the instrumentation tube to maintain physical contact with the interior surface of the instrumentation tube at a plurality of points staggered on a single-diametral plane and spaced substantially throughout the length of the instrumentation tube. The staggered mechanical elements induce a controlled elastic sinuous deflection of the inner thimble tube. The mechanical elements can take several different forms, for example, dimples or cantilevered spring fingers formed in the wall of the instrumentation tube and projecting radially inwardly therefrom.

Abstract: A rod drive control system for nuclear such as boiling water reactors is comprised of a plurality of stepping motors for inserting and withdrawing control rods disposed within the nuclear reactor, which the stepping motors are each provided for each control rod, terminal units which are each provided for each stepping motor and are each provided with a motor controller to drive the related stepping motor and to check the operation of the motor, a central processing unit for transferring an operation signal to the motor control unit and for detecting faults in the terminal units, and transmission lines between the terminal units and the central processing unit.

Abstract: Method and apparatus for monitoring the movement of the control rods of a nuclear reactor.The number of steps of movement in either direction of the rod from which the control rod is suspended is counted. According to the height of the step, an indication of the position of the suspension rod and of the control rod. The apparatus comprises devices for measuring the speed of movement of the control rod, for logging variations in speed higher than a given value, and for counting such variations according to their sign. The invention is particularly useful in pressurized water nuclear reactors.