Abstract: A nuclear steam supply system having a shutdown system for removing residual decay heat generated by a nuclear fuel core. The steam supply system may utilize gravity-driven primary coolant circulation through hydraulic-ally interconnected reactor and steam generating vessels forming the steam supply system. The shutdown system may comprise primary and secondary coolant systems. The primary coolant cooling system may include a jet pump comprising an injection nozzle disposed inside the steam generating vessel A portion of the circulating primary coolant is extracted, pressurized and returned to the steam generating vessel to induce coolant circulation under reactor shutdown conditions. The extracted primary coolant may further be cooled before return to the steam generating vessel in some operating modes. The secondary coolant cooling system includes a pumped and cooled flow circuit operating to circulate and cool the secondary coolant, which in tun extracts heat from and cools the primary coolant.

Abstract: A control rod assembly including at least one movable control rod including a neutron absorbing material, a control rod drive mechanism (CRDM) for controlling movement of the at least one control rod, and a coupling operatively connecting the at least one control rod and the CRDM. The coupling includes a terminal element engaged with a connecting rod of the CRDM and the at least one moveable control rod, and a kinetic energy absorbing element supported by the terminal element for absorbing kinetic energy during a SCRAM event, the kinetic energy absorbing element configured to act between the terminal element and an upper plate of an associated fuel assembly.

Abstract: A nuclear steam supply system having a start-up sub-system for heating a primary coolant. In one embodiment, the invention can be a nuclear steam supply system comprising: a reactor vessel having an internal cavity, a reactor core comprising nuclear fuel disposed within the internal cavity; a steam generating vessel fluidly coupled to the reactor vessel; a primary coolant loop formed within the reactor vessel and the steam generating vessel, a primary coolant in the primary coolant loop; and a start-up sub-system fluidly coupled to the primary coolant loop, the start-up sub-system configured to: (1) receive a portion of the primary coolant from the primary coolant loop; (2) heat the portion of the primary coolant to form a heated portion of the primary coolant; and (3) inject the heated portion of the primary coolant into the primary coolant loop.

Abstract: A power distribution plate (PDP) sits on top of a support plate. Control rod drive mechanism (CRDM) units are mounted on top of the PDP, but the PDP is incapable of supporting the weight of the CRDM units and instead transfers the load to a support plate. The PDP has receptacles which receive cable modules each including mineral insulated (MI) cables, the MI cables being connected with the CRDM units. The PDP may further include a set of hydraulic lines underlying the cable modules and connected with the CRDM units. The cable modules in their receptacles define conduits or raceways for their MI cables and for any underlying hydraulic lines.

Abstract: A control rod guide frame has a central passage of constant cross-section as a function of position along a central axis that passes through the central passage. The central passage is sized and shaped to guide a traveling assembly including at least one control rod as it moves along the central axis. The control rod guide frame comprises at least two radial guide frame sections secured around and defining the central passage. Each radial guide frame section may comprise an extruded radial guide frame section, which may be made of extruded steel. The central passage may include control rod guidance channels parallel the central axis and machined into the extruded radial guide frame sections. The at least two radial guide frame sections may be interchangeable. In some embodiments the at least two radial guide frame sections consist of between four and eight radial guide frame sections.

Abstract: Embodiments are directed to comparison-based methods of conditionally assessing the excess in correlation of an unknown neutron count measurement compared to the correlation present in a data defined as background, and to providing a technical definition of excess correlation intended to properly handle the measured excess correlation. The degree of correlation between an unknown source and background can be used to prevent masking of neutron count data for the source by background radiation.

Abstract: In one embodiment, the steam separation system includes a plurality of steam separators configured to separate liquid from a gas-liquid two-phase flow stream; and a steam dryer system disposed above the steam separators. The steam dryer system includes a plurality of steam dryer bank assemblies and at least one moisture trap. Each of the steam dryer bank assemblies is configured to separate liquid of the gas-liquid two-phase flow stream exiting the steam separators. The moisture trap is disposed next to one of the steam dryer bank assemblies and is configured to remove liquid from the gas-liquid two-phase flow stream.

Abstract: A pressurized water reactor (PWR) includes a vertical cylindrical pressure vessel having a lower portion containing a nuclear reactor core and a vessel head defining an internal pressurizer. A reactor coolant pump (RCP) mounted on the vessel head includes an impeller inside the pressure vessel, a pump motor outside the pressure vessel, and a vertical drive shaft connecting the motor and impeller. The drive shaft does not pass through the internal pressurizer. A central riser may be disposed concentrically inside the pressure vessel, and the RCP impels primary coolant downward into a downcomer annulus between the central riser and the pressure vessel. A steam generator may be disposed in the downcomer annulus and spaced apart from with the impeller by an outlet plenum. A manway may access the outlet plenum so tube plugging can be performed on the steam generator via access through the manway without removing the RCP.

Abstract: The invention relates to an arrangement for producing a proton beam. This arrangement is characterized in that it is constituted by a laser driven accelerator of protons adapted to produce a beam of relativistic protons of 0.5 GeV to 1 GeV with a current in the order of tens of mA, such as a current of 20 mA. The invention can be used for transmutating nuclear waste.

Abstract: A uranium dioxide nuclear fuel pellet has about 50 to about 400 ?M (with respect to a 3-dimentional size) microcells formed of a ceramic material having a chemical attraction with fission products generated in the nuclear fuel pellet to absorb and trap the fission products, such that the extraction of the fission product may be retrained in a normal operation condition and that the performance of the nuclear fuel may be enhanced by mitigating PCI. In addition, highly radioactive fission products including Cs and I having a large generation amount or a long half-life enough to affect the environments can be trapped in the pellet in an accident condition, without being released outside.

Abstract: Disclosed herein is a water-air combined passive feed water cooling apparatus including a water cooling heat exchanger connected to the inside of a containment building to cool down heat of a steam generator using a water cooling method, a cooling tank including the water cooling heat exchanger therein and storing cooling water condensing main steam generated by the steam generator, an evaporative steam pipe connected to the cooling tank, the evaporative steam pipe, into which steam of the cooling water generated by the water cooling heat exchanger in the cooling tank flows, an air cooling heat exchanger connected to the evaporative steam pipe and cooling down and liquefying the steam flowing into the evaporative steam pipe, and a condensed water collecting pipe for refilling the cooling tank with the steam liquefied by the air cooling heat exchanger.

Abstract: A pressurized water nuclear reactor (PWNR) includes a core having a containment shield surrounding a reactor vessel having fuel assemblies that contain fuel rods filled with fuel pellets, and control rods, and a steam generator thermally coupled to the reactor vessel. A flow loop includes the steam generator, a turbine, and a condenser, and a pump for circulating a water-based heat transfer fluid in the loop. The heat transfer fluid includes a plurality of nanoparticles having at least one carbon allotrope or related carbon material dispersed therein, such as diamond nanoparticles.

Abstract: A nozzle repair method and a nuclear reactor vessel include: removing a trepanning portion (208) as a connection portion with respect to an in-core instrumentation cylinder (204) in a groove-welding portion (206); removing the in-core instrumentation cylinder (204) from a lower end plate (66); forming a plug attachment opening (211) by removing the groove-welding portion (206); applying a pressing load to the lower end plate (66) by attaching a plug (212) to the plug attachment opening (211); and welding and fixing the plug (212) attached to the plug attachment opening (211). Accordingly, since a repair is easily performed, it is possible to improve the workability and to decrease a repair cost.

Abstract: A nuclear reactor comprises a nuclear reactor core disposed in a pressure vessel. An isolation valve protects a penetration through the pressure vessel. The isolation valve comprises: a mounting flange connecting with a mating flange of the pressure vessel; a valve seat formed into the mounting flange; and a valve member movable between an open position and a closed position sealing against the valve seat. The valve member is disposed inside the mounting flange or inside the mating flange of the pressure vessel. A biasing member operatively connects to the valve member to bias the valve member towards the open position. The bias keeps the valve member in the open position except when a differential fluid pressure across the isolation valve and directed outward from the pressure vessel exceeds a threshold pressure.

Abstract: A nozzle repair method and a nuclear reactor vessel include: removing a trepanning portion (208) as a connection portion with respect to an in-core instrumentation cylinder (204) in a groove-welding portion (206); removing the in-core instrumentation cylinder (204) from a semi-spherical portion (66) as a lower end plate; forming a surface buttered-welding portion (210) by buttered-welding the surface of the groove-welding portion (206); forming a welding groove (212) by grooving the surface buttered-welding portion (210); inserting a new in-core instrumentation cylinder (204A) provided with a circumferential groove portion (204f) outside an instrumentation equipment guide passage (204d) into an attachment hole (203); and fixing the new in-core instrumentation cylinder (204A) by groove-welding the welding groove (212). Accordingly, since the nozzle welding area is suppressed to a predetermined range, the workability of the repair is improved.

Abstract: A management equipment transfer apparatus which transfers management equipment, which tests and maintains an inside of a nuclear reactor vessel, inside the nuclear reactor vessel while being locked to the management equipment and the nuclear reactor vessel, includes: a cross beam configured to be lengthily disposed inside the nuclear reactor vessel in a transverse direction; a rod configured to be connected to a cross beam lengthily disposed inside the nuclear reactor vessel in a longitudinal direction; a bracket configured to be mounted in the rod to fix the management equipment; an arm configured to extend toward an inner peripheral surface of the nuclear reactor vessel and be connected to the cross beam; and rolling parts configured to be disposed at ends of the arm and the cross beam, contact the inner peripheral surface of the nuclear reactor vessel, and rotatably support the cross beam and the arm.

Abstract: An emergency core cooling system is provided with at least four active safety divisions each equipped with a motor-driven active safety system, and at least one passive safety division equipped with passive system that does not require to be electrically driven. The number of active safety divisions is grater than the number of active safety divisions needed during a design basis accident by two or more, and each active safety division is provided with one motor-driven active safety system. The passive safety system can cool the reactor core without being re-supplied with cooling water from the outside during the time period needed for the active safety system subjected to online maintenance to recover if an accident occurred during online maintenance of one active safety system.

Abstract: A neutron detector system and method for discriminating fissile material from non-fissile material wherein a digital data acquisition unit collects data at high rate, and in real-time processes large volumes of data directly into information that a first responder can use to discriminate materials. The system comprises counting neutrons from the unknown source and detecting excess grouped neutrons to identify fission in the unknown source. Comparison of the observed neutron count distribution with a Poisson distribution is performed to distinguish fissile material from non-fissile material.

Abstract: An apparatus for generating a pressure wave in a liquid medium is disclosed. The apparatus includes a plurality of pressure wave generators having respective moveable pistons, the pistons having respective control rods connected thereto. The apparatus also includes a plurality of transducers coupled to the liquid medium and means for causing the pistons of respective ones of the plurality of the pressure wave generators to be accelerated toward respective ones of the plurality of transducers. The apparatus further includes means for causing restraining forces to be applied to respective control rods to cause respective pistons to impact respective transducers at respective desired times and with respective desired amounts of kinetic energy such that the respective desired amounts of kinetic energy are converted into a pressure wave in the liquid medium.

Abstract: A safety system for a nuclear power plant includes first through fourth sensors; a first division, including a first calculation module that determines first and second calculation results based on signals from the first and second sensors, a first data-sharing module for sharing the first and second calculation results with a second division, and a first voting logic for generating a first safety demand signal based on the first through fourth calculation results; and the second division, including a second calculation module for determining the third and fourth calculation results based on signals from the third and fourth sensors, a second data-sharing module for sharing the third and fourth calculation results with the first division, and a second voting logic for generating a second safety demand signal based on the first, second, third, and fourth calculation results, wherein the first through fourth sensors each monitor the same plant parameters.