Abstract: To stabilize boiling water reactors against neutron flux oscillations caused by thermal-hydraulic instabilities, random core-wide reactivity perturbations are induced in the reactor either to suppress the onset of asymmetric (regional) mode oscillations or, in response to the detection of asymmetric mode oscillations, to stabilize the reactor in a symmetric (core-wide) oscillation mode. The reactivity perturbations are produced by effecting random changes in dome pressure, core flow or coolant enthalpy.

Abstract: An apparatus for refurbishing the ends of control rod drive (CRD) penetrations used in nuclear boiling water reactor systems where the rods are disposed in a grid of rows and columns with a known predetermined spacing therebetween. The apparatus is rapidly positioned for support on a pair of CRDs adjacent to a CRD to be refurbished, coaxially aligning the end mill with the CRD to be refurbished. The end mill is then brought into contact with the CRD to be refurbished. After the refurbishment process, the apparatus is rapidly removed and ready to be positioned on another pair of CRDs adjacent to another CRD to be refubished.

Abstract: A method of avoiding localized hydrogen build-ups in the atmospheres of safety tanks of reactors, especially light water reactors, is provided. After a break down that is accompanied by loss of coolant has occurred, the cooling water temperature in the sump of the safety tank is adjusted in the phase of the long term cooling to a temperature that is high than the air temperature in the dome of the safety tank, with this temperature adjustment being effected in such a way that an adequate intermixing of the atmosphere in said safety tank is effective in a convective manner.

Abstract: In 1974, a well-known research problem in Statistical Mechanics entitled "To determine and define the probability function P.sub.2 of a particle hitting a predesignated area, given all its parameters of generation and ejection." was openly solicited for its solution from research and development organizations in U.S.A. One of many proposed solutions of the problem, initiated at that time, is by means of the TRAJECTORY SOLID ANGLE(TSA). TSA is defined as the integral of the dot product of the unit tangent of the particle's trajectory to the vectorial area divided by the square of the position vector connecting between the point of ejection and that of the surface to be hit. The invention provides:(1) The precise and the unique solution of a previously unsolved P.sub.

Abstract: A reactor system of a boiling water type atomic power plant including a reactor pressure vessel, a primary containment vessel containing the reactor pressure vessel, a main turbine, a main steam piping extending through the primary containment vessel between the reactor pressure vessel and the main turbine, and inside and outside main steam isolation valves provided on the main steam piping inside and outside the primary containment vessel, respectively. A closure valve is provided on the main steam piping at a position near the primary containment vessel and is adapted to quickly close in less than 3 seconds upon a break accident of the main steam piping. A volume of the reactor pressure vessel allotted for accommodating steam is correspondingly enlarged to allow the closure valve to quickly close upon a break accident of the main steam piping.

Abstract: An end cap is mounted on a fuel rod gripper to cover the end portions of fingers thereon to minimize impacts with grid cell dimples. The fingers are defined by circumferentially-spaced and longitudinally-extending slots in the gripper which permit the finger end portions to radially expand and retract between corresponding expanded and collapsed conditions. The end cap includes a body having a forward camming portion and a rearward mounting portion. The rearward mounting portion is hollow and configured to fit over and cover the end portions of the gripper fingers with the forward camming portion extending forwardly thereof. Also, circumferentially-spaced and longitudinally-extending attachment flutes are disposed internally of the hollow mounting portion of the body and project radially inwardly therefrom.

Abstract: A fluid pressure actuated steam depressurization valve for nuclear reactor service. The valve is maintained closed retaining steam pressure within the system by the application of fluid pressure from an external source and control, and upon release of the applied fluid pressure steam pressure within the nuclear reactor system is vented to depressurize the system.

Abstract: Protection against the consequences of a cluster of control rods falling accidentally is provided by four chains, each of which includes two neutron flux detectors (D1A, D1B), two primary treatment circuits (R1A, S1A and R1B, S1B) associated with said two detectors such that each provides a primary fall signal in the event of the flux measured by the corresponding detector decreasing rapidly, and an OR gate (P1) for providing an output fall signal from the chain in the event of at least one of said primary signals being present. The detectors (D1A, D4B, D3A, D1B, D2A, D3B, D4A and D2B) are angularly distributed around the vertical axis (A) of the core (2). A secondary treatment circuit (6) causes an emergency reactor stop to take place in the event that it receives fall signals at the outputs from at least two chains.

Abstract: A member adapted for use under a stress in an atmosphere of a temperature below the creep temperature and made from an Ni base alloy having a high resistance to stress corrosion cracking. The Ni alloy consists essentially of, by weight, 15 to 25% of Cr, 1 to 8% of Mo, 0.4 to 2% of Al, 0.7 to 3% of Ti, 0.7 to 4.5% of Nb and the balance Ni, and has an austenite matrix in which precipitated is at least one of .gamma.' phase and .gamma." phase. The member can suitably used as parts which are subjected to pure water in nuclear reactor.

Abstract: An electrical ground fault detection and limitation system for employment with a nuclear reactor utilizing a liquid metal coolant. Elongate electromagnetic pumps submerged within the liquid metal coolant and electrical support equipment experiencing an insulation breakdown occasion the development of electrical ground fault current. Without some form of detection and control, these currents may build to damaging power levels to expose the pump drive components to liquid metal coolant such as sodium with resultant undesirable secondary effects. Such electrical ground fault currents are detected and controlled through the employment of an isolated power input to the pumps and with the use of a ground fault control conductor providing a direct return path from the affected components to the power source.

Abstract: A method and a device are disclosed for measuring the lithium concentration in the primary cooling circuit of a nuclear power station reactor. The lithium concentration Li.sup.+ is determined by linear correlation between the electrical conductivity .lambda. of the cooling water measured by measuring means 1, at a given temperature t measured by measuring means 2. Means 4 to measure the boron concentration B make it possible, by means of the computing means 3 to determine the lithium concentration Li.sup.+ according to the decreasing lithium hydroxide specification. The invention can be applied to the maintaining and regulation of the pH of the cooling water of a nuclear power station reactor at a determined value.

Abstract: A nuclear reactor includes a reactor vessel, a reactor core disposed within the vessel and including fuel rods, a primary cooling system disposed in the vessel and including a steam generator and a pump and a poison tank disposed within the vessel and containing high concentration boric acid solution. The poison tank surrounds the reactor core, and the steam generator surrounds a portion of the poison tank.

Abstract: Device for preventing thermal stratification for a steam generator feed pipe. A fixed helix (14) downstream of a rising break (12) of the pipe subdivides its interior into twisted channels. The cold water entering at a low flame rate is blocked upstream until a predetermined height (h) is reached and completely fills an upstream part (11) to be protected against thermal gradients.

Abstract: A nuclear power plant with a gas cooled high temperature reactor, installed in a prestressed concrete pressure vessel with the operational and decay heat removal systems. The prestressed concrete pressure vessel has a thermal protection system, with a thermal insulating layer and a liner cooling system. The liner cooling system, which is includes water carrying cooling pipes, which along with intermediate heat exchangers and cooling water pumps make up a closed intermediate cooling loop used for removal of the decay heat in case a failure of the decay heat removal systems. The elements of the invention assure an adequate water flow for the removal of decay heat in the liner cooling system in any situation, i.e. such that the decay heat may also be removed by natural convection. These element insure a sufficient driving pressure differences and minimize pressure losses in the intermediate cooling loop.

Abstract: A nuclear reactor includes a plurality of upstanding guide thimbles and a plurality of control rods received in the guide thimbles and supported for movement relative to the thimbles between inserted and withdrawn positions. The control rods each include a tubular cladding member and an end plug attached to a lower end of the member. The improvement relates to the end plug having a stabilizing configuration which reduces lateral vibratory motion and contact between the control rod and its respective guide thimble. The end plug has an asymmetrical shape which places its terminal tip end in a position offset to one side of the central axis of the control rod and its end plug. As a result of interaction of the asymmetrical plug tip with coolant flowing along the control rod, a lateral steady-state force is imparted to the end plug which maintains the control rod end plug pressed against the wall of tis respective guide thimble. Several different asymmetrical configurations can be employed on the end plug.

Abstract: A pressurized water nuclear reactor has a reactor vessel arranged in a pool, which is filled with a neutron absorbing liquid, for example borated water. The reactor vessel is closed except for tubes connecting it with a tray above it. The coolant in the circuit rises from the vessel to the tray, gives up its heat by flashing, and flows back to the bottom of the vessel, driven by natural circulation. The tray is separated from the pool by a vapor-filled bell, which surrounds it. In the bell the vapor gives up its useful heat to a condenser. The relatively low boron content of the cooling circuit, compared to the pool, is achieved by continuous dilution of the condensate from vapor additionally generated out of the pool water. The dilution process is an equilibrium with continuous inflow of the pool water. The inflow is automatically controlled by the pool level, which rises when the pool water is pressed out from below the bell by overproduction of vapor.

Abstract: A nuclear steam generator isolation dam (28) locking pin (36) is slidably mounted and spring (52) biased in an axial bore (42,44) of a body (40) press fit in a dam boss (32). The pin (36) is actuated by a lever (54) in a slot (54,58,62) for selective lock-in, lock-out and spring biased travel for seeking its keeper (27) in the wall (26) of the nozzle.

Abstract: A bolt and keeper secure nuclear steam generation isolation dam segments in assembled relation. The bolt is rotated into locked position with its handle in a slot as a transverse projection on the bolt end is moved along a ramp into a securing notch against the bias of a compressible rubber spring.

Abstract: A stuck fuel rod capping sleeve to be used during derodding of spent fuel assemblies if a fuel rod becomes stuck in a partially withdrawn position and, thus, has to be severed. The capping sleeve has an inner sleeve made of a lower work hardening highly ductile material (e.g., Inconel 600) and an outer sleeve made of a moderately ductile material (e.g., 304 stainless steel). The inner sleeve may be made of an epoxy filler. The capping sleeve is placed on a fuel rod which is then severed by using a bolt cutter device.

Abstract: A steam control value for a steam chest such as used in a nuclear plant with an improved muffler portion. The muffler portion includes window openings of predetermined dimensions. A ring portion is below the window opening having a plurality of pockets. Each pocket is aligned with one of the window openings. This design greatly reduces vibrations which may damage the valve.

Abstract: A nuclear reactor vessel has an electromagnet positioned within to generate the predetermined temperature for the vessel wall for a predetermined time to anneal the vessel.

Abstract: The invention relates to a standby or emergency cooling device for the core of a pressurized water reactor.A high pressure reservoir filled with boric acid solution is permanently connected to the reactor vessel via two pipes, preferably arriving at the same level. However, one of these pipes is extended within the vessel by a downwardly directed bend in order to issue below the other pipe. These two pipes have horizontal portions which are intended to prevent any natural circulation under normal operating conditions.Application to the improvement of the safety of pressurized water nuclear reactors.

Abstract: In a method of inhibiting or avoiding intermixing of the atmosphere which exists in a closed nuclear reactor containment space with a gaseous substance present in the same space a change in density of the atmosphere is induced in either lower or upper area of the space such that a stable atmosphere stratified into superimposed layers is formed in that space due to the change in density.

Abstract: A steam generator for transferring the heat carried by a liquid metal to a water - steam circuit. It comprises a drainage tank directly suspended at the lower end of the envelope. A drainage tube connected to the bottom of the envelope is disposed within the tank and carries a burster disk at its end. This disk is placed behind a detachable observation window, fixed to the drainage tank and permitting the replacement thereof.

Abstract: A self-test system for a nuclear power plant, nuclear reactor protection system is disclosed. Nuclear protection systems are the electronic controls, typically including circuit cards, located intermediate between sensors (as for detecting core overheat) and a control (as for providing rod injection to shut down a reactor). Constant surveillance of the nuclear system protection system is provided by a microprocessor that serially addresses protection system circuit cards and loads them at pre-determined input points with test commands. The addressed cards are thereafter simultaneously activated by a system-wide command. The test command is a pulse which is so short in duration that its affect is transparent to the system and cannot cause overall system operation. The pulse passes through the actuating electrical components to verify, on the real actuating path, the operating integrity of the system.

Abstract: It is desirable to being able to vary the temperature of a loop exposed to a flow of hot gases in gas cooled nuclear reactor installations without thereby affecting the gas temperature in the other loops. It should be possible in an extreme case to take the affected loop out of service. This is achieved by arranging under the heat absorbing components in the hot gas channel a means for mixing gas connected with a cold gas line.

Abstract: Fuel assemblies of a nuclear reactor are transferred during fueling or refueling or the like by a crane. The work-engaging fixture of the crane picks up an assembly, removes it from this slot, transfers it to the deposit site and deposits it in its slot at the deposit site. The control for the crane includes a strain gauge connected to the crane line which raises and lowers the load. The strain gauge senses the load on the crane. The signal from the strain gauge is compared with setpoints; a high-level setpoint, a low-level setpoint and a slack-line setpoint. If the strain gauge signal exceeds the high-level setpoint, the line drive is disabled. This event may occur during raising of a fuel assembly which encounters resistance. The high-level setpoint may be overridden under proper precautions. The line drive is also disabled if the strain gauge signal is less than the low-level setpoint.

Abstract: A method and apparatus is shown for establishing a predetermined concentration of gas in the reactor core coolant for a pressurized water nuclear power system. Gas is added to core coolant water in a bubble column. The gas-water mixture flows through static mixers in the column to produce the desired gas concentration in the water.

Abstract: The invention relates to an emergency shut-off device, in the case of leakage, of a tube of a steam generator where the heat of vaporization is contributed by a liquid metal circulating in contact with the outer surface of the tube. The device includes on the terminal portion of the tube connected to the water supply device a tubular bush interposed in the path of the feed water. The bush contains a valve connected to the bush by a rupture part. The breakage of the part under the effect of an increased flow of water, in the case of a leakage, enables the closing of the outlet of the bush. The device includes also, in the terminal portion of the tube connected to the steam collector a bush containing a free valve enabling the steam outlet to be closed in the case of a leakage in the tube. The invention is applied in particular to steam generators of fast breeder nuclear reactors cooled by liquid sodium.

Abstract: A jet pump plug for a multi-nozzle jet pump is disclosed wherein the nozzles are accessible through a relatively small opening in the pump structure. The plug comprises a plurality of sections, each including a closure plate and a lever coupled thereto. Each plug section is constructed so as to allow that section to be separately introduced into the opening to the vicinity of the nozzles from a location which may be remote from the pump. The closure plates are manipulated through their attached levers from a remote location to the end of sealing the nozzles with them. The first plug section introduced through the opening includes means for alignment relative to the jet pump structure so as to orient its closure plate with respect to the nozzles. The remaining plug section is supported by the first plug section upon being introduced into the opening.

Abstract: The invention relates to a protective device for the actuating mechanisms of the control clusters during tests under pressure of the primary circuit of a nuclear reactor. Before mounting the clusters, each guidance sheath for the parts connecting the clusters with the control mechanisms is provided with a removable filter dismountably fixed at the lower portion of the sheath. The filter is constituted by several metal gauzes. The invention is applicable to pressurized water nuclear reactors.

Abstract: A fluid circulation system for heat exchangers having two groups of tubes through which primary and secondary fluids flow, the tubes of one group being interdigitated with the tubes of the other group, and a heat transfer material interposed between the two groups of tubes, whereby heat is transferred from the primary fluid through the heat transfer material to the secondary fluid. A shell forms a closure around the tubes and the heat transfer material, and the shell has tertiary fluid inlet and outlet means. Openings in the heat transfer material form passageways through which the tertiary fluid can flow from the inlet means, through the heat exchanger, to the outlet means. Piping connects the tertiary fluid outlet means to the tertiary fluid inlet means, forming a complete cycle. Installed in this piping is a heat removal system. If the secondary fluid flow is interrupted, the tertiary fluid provides a redundant means by which the heat of the primary fluid can be removed from the heat exchanger.

Abstract: A nuclear power plant comprising a control building for receiving therein control instruments which make controls of the nuclear energy plant, a reactor building for receiving a nuclear reactor therein, a turbine building for receiving therein turbines and a generator which are driven by steam supplied from the nuclear reactor, said buildings being arrayed in a straight line, and scattering fragments-checking means disposed on the inner wall of a turbine casing, said means functioning so that when a turbine rotary body breaks and scatters, the scattering fragments may collide against it and have their rescattering directions limited.