公开(公告)号：US20130202075A1

公开(公告)日：2013-08-08

申请号：US13614898

申请日：2012-09-13

申请人： Tae Soon Kwon ,  Chul Hwa Song ,  Joon Eon Yang ,  Won Pil Baek

发明人： Tae Soon Kwon ,  Chul Hwa Song ,  Joon Eon Yang ,  Won Pil Baek

IPC分类号： G21C9/00

CPC分类号： G21C15/18 ,  G21C19/28 ,  G21D1/02 ,  Y02E30/40

摘要： A residual heat removal system for a nuclear power plant. The residual heat removal system for a nuclear power plant may include an air duct provided on an outside of a reactor containment building, a heat exchanger disposed on an inside of the air duct, a first pipe to transfer, to the heat exchanger, steam generated in a steam generator disposed on an inside of the reactor containment building, and second pipe to transfer, to the steam generator, water condensation that is cooled and condensed in the heat exchanger, wherein the heat exchanger is air-cooled using outside air flowing inside of the air duct.

摘要翻译： 一个用于核电站的剩余散热系统。 用于核电站的剩余散热系统可以包括设置在反应堆容纳建筑物的外部的空气管道，设置在空气管道内侧的热交换器，将热交换器传送到热交换器的第一管道 在设置在反应堆容纳建筑物的内部的蒸汽发生器中，以及第二管道，用于将热交换器中冷却和冷凝的水冷凝物传递到蒸汽发生器，其中，热交换器使用在内部流动的外部空气进行空气冷却 的风道。

公开(公告)号：US08483349B2

公开(公告)日：2013-07-09

申请号：US12181891

申请日：2008-07-29

申请人： Jae-Yong Kim ,  Hyung-Kyu Kim ,  Kyung-Ho Yoon ,  Young-Ho Lee ,  Kang-Hee Lee

发明人： Jae-Yong Kim ,  Hyung-Kyu Kim ,  Kyung-Ho Yoon ,  Young-Ho Lee ,  Kang-Hee Lee

IPC分类号： G21C3/34

CPC分类号： G21C3/02 ,  G21C3/322 ,  G21C3/352 ,  G21C3/356 ,  Y02E30/38

摘要： A spacer grid for dual-cooling nuclear fuel rods arranged at a narrow interval. The spacer grid solves the problem in which, since the dual-cooling nuclear fuel rods are used to improve the cooling performance and stability of nuclear fuel and obtain high burnup and output, the outer diameter of each dual-cooling nuclear fuel rod is increased, and thus the gap between each dual-cooling nuclear fuel rod and the grid strap is decreased. The spacer grid includes first grid straps and second grid straps, which are crossed and arranged in transverse and longitudinal directions at regular intervals and have the shape of a flat strip, and support structures, which are fitted into the first and second grid straps around intersections of the first and second grid straps so as to support the dual-cooling nuclear fuel rods.

摘要翻译： 用于双冷却核燃料棒的隔离栅格，其间隔狭窄。 间隔栅格解决了由于双冷核燃料棒用于提高核燃料的冷却性能和稳定性并获得高燃耗和输出的问题，每个双重冷却核燃料棒的外径增加， 因此，每个双重冷却核燃料棒和网格带之间的间隙减小。 间隔网格包括第一网格带和第二网格带，它们以横向和纵向方向以规则的间隔交叉并布置，并且具有平坦的形状和支撑结构，其被安装在交叉点周​​围的第一和第二网格带中 的第一和第二格栅带，以支撑双重冷却核燃料棒。

公开(公告)号：US20130070887A1

公开(公告)日：2013-03-21

申请号：US13615779

申请日：2012-09-14

申请人： Young In Kim ,  Young Dong Hwang ,  Young Jong Chung ,  Won Jae Lee ,  Keung Koo Kim ,  Hwang Bae ,  Jung Yoon ,  Sung Kyun Zee ,  Cheon Tae Park

发明人： Young In Kim ,  Young Dong Hwang ,  Young Jong Chung ,  Won Jae Lee ,  Keung Koo Kim ,  Hwang Bae ,  Jung Yoon ,  Sung Kyun Zee ,  Cheon Tae Park

IPC分类号： G21C9/00

CPC分类号： G21C15/18 ,  G21C17/022 ,  Y02E30/40

摘要： Disclosed are a nuclear reactor adapted for mitigating a loss-of-coolant accident and a mitigation method thereof. The nuclear reactor includes a nuclear reactor vessel, a first pipe part connected to the nuclear reactor vessel to supply or drain fluid, and a first opening/closing part connected to the first pipe part. When the first pipe part is broken, the first opening/closing part closes the first pipe part to stop discharge of coolant. A supplementary coolant injection part is connected to the nuclear reactor vessel through a second pipe part. When the second pipe part is broken, a second opening/closing part closes the second pipe part to stop discharge of coolant.

摘要翻译： 公开了一种适用于减轻冷却剂损失事故的核反应堆及其缓解方法。 核反应堆包括核反应堆容器，连接到核反应堆容器以供应或排出流体的第一管部分和连接到第一管部分的第一开/关部分。 当第一管部分断裂时，第一开/关部分关闭第一管部分以停止冷却剂的排放。 辅助冷却剂喷射部分通过第二管部连接到核反应堆容器。 当第二管部分断裂时，第二开/闭部分关闭第二管部分以停止冷却剂的排出。

公开(公告)号：US08381803B2

公开(公告)日：2013-02-26

申请号：US12515361

申请日：2007-10-10

申请人： Youngwan Kim ,  Jaewon Park ,  Minhwan Kim ,  Sungdeok Hong ,  Wonjae Lee ,  Jonghwa Chang

发明人： Youngwan Kim ,  Jaewon Park ,  Minhwan Kim ,  Sungdeok Hong ,  Wonjae Lee ,  Jonghwa Chang

IPC分类号： F28F19/02 ,  F28F3/02

CPC分类号： G21D9/00 ,  C01B3/06 ,  F28D9/0062 ,  F28D15/00 ,  F28F3/025 ,  F28F19/02 ,  F28F21/08 ,  Y02E60/364

摘要： A high-temperature and high-pressure corrosion-resistant process heat exchanger for a nuclear hydrogen production system decomposes sulfite (SO3) using heat from a high-temperature gas-cooled reactor to thereby produce sulfide (SO2) and oxygen (O2). The process heat exchanger comprises second and third system coolant channels, each of which is defined by a heat transmission fin, which is bent in a quadrilateral shape, and heat transmission plates, and has increased corrosion resistance thanks to ion-beam coating and ion-beam mixing using a material having high corrosion resistance. The third system coolant channel includes reaction catalysts for SO3 decomposition, and is made of a super alloy. Thus, a system differential pressure between the second and third system coolant channels can be greatly maintained at a high temperature of 900° C. or higher.

摘要翻译： 用于核氢生产系统的高温高压耐腐蚀工艺热交换器利用高温气冷反应堆的热量分解亚硫酸盐（SO3），从而产生硫化物（SO2）和氧气（O2）。 过程热交换器包括第二和第三系统冷却剂通道，每个通道由四边形弯曲的传热翅片和传热板限定，并且由于离子束涂覆和离子束涂覆而具有增加的耐腐蚀性， 使用具有高耐腐蚀性的材料进行光束混合。 第三系统冷却剂通道包括用于SO 3分解的反应催化剂，并且由超级合金制成。 因此，可以在900℃以上的高温下极大地维持第二和第三系统冷却剂通道之间的系统压差。

公开(公告)号：US08300759B2

公开(公告)日：2012-10-30

申请号：US12355088

申请日：2009-01-16

申请人： Hae Yong Jeong ,  Chungho Cho ,  Yong Bum Lee ,  Dong Uk Lee ,  Jae Hyuk Eoh ,  Kwi Seok Ha

发明人： Hae Yong Jeong ,  Chungho Cho ,  Yong Bum Lee ,  Dong Uk Lee ,  Jae Hyuk Eoh ,  Kwi Seok Ha

IPC分类号： G21C9/00

CPC分类号： G21C15/18 ,  Y02E30/40

摘要： Disclosed herein is a decay heat removal system, including: a decay heat exchanger that absorbs decay heat generated by a nuclear reactor; a heat pipe heat exchanger that receives the decay heat from the decay heat exchanger through a sodium loop for heat removal and then discharges the decay heat to the outside; and a sodium-air heat exchanger that is connected to the heat pipe heat exchanger through the sodium loop and discharges the decay heat transferred thereto through the sodium loop to the outside. According to the decay heat removal system, a heat removal capability can be realized by the heat pipe heat exchanger at such a high temperature at which the safety of a nuclear reactor is under threat, and a cooling effect can be obtained through the sodium-air heat exchanger at a temperature lower than that temperature.

摘要翻译： 本文公开了一种衰变除热系统，包括：衰变热交换器，其吸收由核反应堆产生的衰变热; 热管式热交换器，其通过钠回路接收来自衰变热交换器的衰减热，用于除热，然后将衰变热量排出到外部; 以及钠 - 空气热交换器，其通过钠回路连接到热管热交换器，并将通过钠回路传递到其的衰变热排出到外部。 根据衰变除热系统，在核反应堆的安全性受到威胁的高温下，热管换热器可以实现除热能力，通过钠空气可以获得冷却效果 热交换器的温度低于该温度。

公开(公告)号：US08268204B2

公开(公告)日：2012-09-18

申请号：US12552480

申请日：2009-09-02

申请人： Young Woo Rhee ,  Dong Joo Kim ,  Kun Woo Song ,  Ki Won Kang ,  Jong Hun Kim ,  Keon Sik Kim ,  Jae Ho Yang

发明人： Young Woo Rhee ,  Dong Joo Kim ,  Kun Woo Song ,  Ki Won Kang ,  Jong Hun Kim ,  Keon Sik Kim ,  Jae Ho Yang

IPC分类号： B22F3/12

CPC分类号： B30B11/02 ,  B22F2003/033 ,  B22F2998/00 ,  B22F2998/10 ,  B22F2999/00 ,  B28B3/027 ,  B30B15/022 ,  C04B35/51 ,  C04B2235/3284 ,  C04B2235/604 ,  C04B2235/608 ,  C04B2235/652 ,  C04B2235/6582 ,  C04B2235/77 ,  C04B2235/94 ,  C04B2235/9638 ,  G21C21/02 ,  G21C2003/045 ,  B22F5/106 ,  B22F1/007 ,  B22F3/02 ,  B22F3/1007 ,  B22F2201/013 ,  B22F2201/04 ,  B22F2201/05 ,  B22F2201/10

摘要： A method for manufacturing an annular nuclear fuel pellet is provided. In the method, an annular nuclear fuel green compact whose lateral cross-section is a trapezoid is prepared. The thickness of the annular nuclear fuel green compact reduces along one direction of the central axis, and a green density of the nuclear fuel green compact increases along one direction of the central axis. The annular nuclear fuel green compact is sintered under a reducing gas atmosphere so that the annular nuclear fuel pellet is obtained. According to this method, the annular pellet which has uniform inner and outer diameters and small diametric tolerances along the pellet height is fabricated without grinding the pellet surfaces.

摘要翻译： 提供一种制造环形核燃料颗粒的方法。 在该方法中，准备了横截面为梯形的环状核燃料生坯。 环形核燃料生坯的厚度沿着中心轴的一个方向减小，并且核燃料生坯的绿色密度沿着中心轴的一个方向增加。 环形核燃料生坯在还原气氛下烧结，得到环状核燃料颗粒。 根据该方法，在不研磨颗粒表面的情况下制造具有均匀的内径和外径均匀且小颗粒高度的小直径公差的环形颗粒。

公开(公告)号：US08243872B2

公开(公告)日：2012-08-14

申请号：US12146493

申请日：2008-06-26

申请人： Kang-Hee Lee ,  Hyung-Kyu Kim ,  Kyung-Ho Yoon ,  Young-Ho Lee ,  Jae-Yong Kim ,  Tae-Hyun Chun ,  Wang-Kee In ,  Dong-Seok Oh ,  Chang-Hwan Shin

发明人： Kang-Hee Lee ,  Hyung-Kyu Kim ,  Kyung-Ho Yoon ,  Young-Ho Lee ,  Jae-Yong Kim ,  Tae-Hyun Chun ,  Wang-Kee In ,  Dong-Seok Oh ,  Chang-Hwan Shin

IPC分类号： G21C3/34

CPC分类号： G21C3/352 ,  G21C3/322 ,  G21C3/3563 ,  Y02E30/38

摘要： A spacer grid can be applied to close-spaced nuclear fuel rods. The spacer grid is directed to solve the problem in which, as the outer diameter of each nuclear fuel rod increases due to the use of dual-cooled nuclear fuel rods for improving cooling performance and obtaining high combustion and high output power, the gap between the neighboring nuclear fuel rods is narrowed to thus make it impossible to use an existing spacer grid. The spacer grid is a combination of unit grid straps, each of which has supports for supporting each of the nuclear fuel rods set in a narrow array and has a sheet shape, which are combined with each other. The supports are located at positions shifted from the longitudinal central line of each unit grid strap toward sub-channels.

摘要翻译： 间隔栅格可以应用于紧密间隔的核燃料棒。 针对这种间隔格栅来解决这样一个问题，即由于使用双冷核燃料棒来提高冷却性能并获得高燃烧和高输出功率，因此每个核燃料棒的外径增加 相邻的核燃料棒变窄，因此不可能使用现有的间隔栅格。 间隔栅格是单元网格带的组合，每个网格带都具有支撑，用于支撑以窄阵列设置的每个核燃料棒，并且具有彼此组合的片状。 支撑件位于从每个单元格栅带的纵向中心线向子通道移位的位置。

公开(公告)号：US20120183111A1

公开(公告)日：2012-07-19

申请号：US13325322

申请日：2011-12-14

申请人： Kwon-Hae Yea ,  HyungKi Cha ,  Sung-Man Lee ,  Sang-Soon Park ,  Seong Hee Park

发明人： Kwon-Hae Yea ,  HyungKi Cha ,  Sung-Man Lee ,  Sang-Soon Park ,  Seong Hee Park

IPC分类号： G21G1/06

CPC分类号： G21G1/06 ,  G01N23/222 ,  G01N2223/204

摘要： Disclosed herein is a prompt gamma-ray detection apparatus for analyzing chemical materials using femtosecond pulse laser-induced neutrons, which can be effectively used in the nondestructive inspection of various materials, such as metals, coal, cement, radioactive materials and the like as well as explosives and chemical materials, and which can provide better measurement results for the analysis of basic materials, and a method of measuring prompt gamma-rays using the apparatus. The prompt gamma-ray detection apparatus is advantageous because it can non-destructively analyze the elements in a chemical sample using a femtosecond pulse laser-induced neutron generator that solves the problems of an atomic reactor for research or a radioactive isotope as a neutron radiation source.

摘要翻译： 本文公开了一种用于使用飞秒脉冲激光诱导中子分析化学材料的快速γ射线检测装置，其可以有效地用于各种材料如金属，煤，水泥，放射性材料等的非破坏性检查 作为爆炸物和化学材料，并且可以为基础材料的分析提供更好的测量结果，以及使用该装置测量快速γ射线的方法。 迅速的伽马射线检测装置是有利的，因为它可以使用飞秒脉冲激光诱导中子发生器对化学样品中的元素进行非破坏性分析，该中子发生器解决用于研究的原子反应堆或放射性同位素作为中子辐射源的问题 。

公开(公告)号：US20120145287A1

公开(公告)日：2012-06-14

申请号：US12392697

申请日：2009-02-25

申请人： Hyun Gil Kim ,  Yong Hwan Jeong ,  Byoung Kwon Choi ,  Sang Yoon Park ,  Myung Ho Lee ,  Jeong Yong Park ,  Jun Hwan Kim

发明人： Hyun Gil Kim ,  Yong Hwan Jeong ,  Byoung Kwon Choi ,  Sang Yoon Park ,  Myung Ho Lee ,  Jeong Yong Park ,  Jun Hwan Kim

IPC分类号： C22F1/18 ,  C22C16/00

CPC分类号： C22C16/00

摘要： Disclosed herein are a zirconium alloy composition, which exhibits excellent corrosion resistance by varying the kinds of metal oxides and controlling the size of precipitates of the composition, including: 1.05˜1.45 wt % of Nb; one or more selected from the group consisting of 0.1˜0.7 wt % of Fe and 0.05˜0.6 wt % of Cr; and residual Zr, and a method of preparing the same. The zirconium alloy composition exhibits excellent corrosion resistance by controlling the kinds and amounts of the elements included in the zirconium alloy composition and the heat-treatment temperature and thus varying the kinds of metal oxides formed during an oxidation process and controlling the size of precipitates of the zirconium alloy, so that it can be usefully used as a raw material for nuclear fuel cladding tubes, spacer grids, nuclear reactor internals and the like of a light-water reactor or a heavy-water reactor in a nuclear power plant.

摘要翻译： 本发明公开了一种锆合金组合物，其通过改变金属氧化物的种类和控制组合物的析出物的尺寸来表现出优异的耐腐蚀性，包括：1.05〜1.45重量％的Nb; 选自由0.1〜0.7重量％的Fe和0.05〜0.6重量％的Cr组成的组中的一种或多种; 和残余Zr，及其制备方法。 锆合金组合物通过控制包含在锆合金组合物中的元素的种类和数量以及热处理温度，从而改变氧化过程中形成的金属氧化物的种类和控制沉淀物的尺寸，显示出优异的耐腐蚀性 锆合金，因此可以有效地用作核电厂的轻水反应堆或重水反应堆的核燃料包层管，间隔栅格，核反应堆内部等的原料。

公开(公告)号：US20120051978A1

公开(公告)日：2012-03-01

申请号：US13105010

申请日：2011-05-11

申请人： Young Hwan Kim ,  Geun Il Park ,  Jae Hoo Jung ,  Jung Won Lee ,  Ki Ho Kim ,  Young Soon Lee ,  Do Youn Lee

发明人： Young Hwan Kim ,  Geun Il Park ,  Jae Hoo Jung ,  Jung Won Lee ,  Ki Ho Kim ,  Young Soon Lee ,  Do Youn Lee

IPC分类号： G21C1/01

CPC分类号： G21C19/44 ,  Y02W30/882

摘要： A vol-oxidizer for spent nuclear fuel is provided. The vol-oxidizer includes a reactor module in which the spent nuclear fuel is put and oxidized; a heating module to heat the reactor module to a high temperature; a utility module to control an inner state of the reactor module, being connected with the reactor module; a support module to support the heating module and the reactor module; a drive module to supply a driving force for transferring the spent nuclear fuel within the reactor module; a valve module to control discharge of the spent nuclear fuel being oxidized in the reactor module; and a collecting container module to collect the spent nuclear fuel being oxidized and discharged from the reactor module. The respective modules are assembled and disassembled with respect to one another.

摘要翻译： 提供了用于乏核燃料的挥发氧化剂。 体积氧化剂包括反应堆模块，其中废核燃料被放置和氧化; 将反应器模块加热到高温的加热模块; 用于控制反应堆模块的内部状态的实用模块，与反应堆模块连接; 支撑模块，用于支撑加热模块和反应堆模块; 驱动模块，用于提供用于在所述反应堆模块内传送废核燃料的驱动力; 阀模块，用于控制在所述反应堆模块中被氧化的废核燃料的排放; 以及收集容器模块，用于收集从反应器模块氧化和排出的废核燃料。 相应的模块相对于彼此组装和拆卸。