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Survey of experiments and code development for the passive residual heat removal system of pwr in China

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Nuclear Science and Techniques 2004年第1期15卷 53-58页

作者： HUANGYan-Ping ZHUOWen-Bing YANGZu-Mao XIAOZe-Jun CHENBing-De JIADou-Nan Xi＇anJiaotongUniversity Xi＇an710049 NuclearPowerInstituteofChina Chengdu610041

Three different kinds of experiments and their typical results are surveyed for the passive residual heat removal system (PRHRS) of pwr performed in Nuclear Power Institute of China (NPIC) recent ten years. The typi- cal results of MISAP, a special code for pwr passive residual heat removal system developed and assessed by NPIC, are also described briefly in this paper.

关键词：中国无源剩余热清除系统 pwr PRHRS 反应堆核电站

Effect of water injection on hydrogen generation during severe accident in pwr

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Nuclear Science and Techniques 2009年第5期20卷 312-316页

作者： TAO Jun CAO Xuewu School of Mechanical and Power Engineering Shanghai Jiao Tong University Shanghai 200240 China

Effect of water injection on hydrogen generation during severe accident in a 1000 MWe pressurized water reactor was studied. The analyses were carried out with different water injection rates at different core damage stages. The core can be quenched and accident progression can be terminated by water injection at the time before cohesive core debris is formed at lower core region. Hydrogen generation rate decreases with water injection into the core at the peak core temperature of 1700 K, because the core is quenched and reflooded quickly. The water injection at the peak core temperature of 1900 K, the hydrogen generation rate increases at low injection rates of the water, as the core is quenched slowly and the core remains in uncovered condition at high temperatures for a longer time than the situation of high injection rate. At peak core temperature of 2100–2300 K, the Hydrogen generation rate increases by water injection because of the steam serving to the high temperature steam-starved core. Hydrogen generation rate increases significantly after water injection into the core at peak core temperature of 2500 K because of the steam serving to the relocating Zr-U-O mixture. Almost no hydrogen generation can be seen in base case after formation of the molten pool at the lower core region. However, hydrogen is generated if water is injected into the molten pool, because steam serves to the crust supporting the molten pool. Reactor coolant system (RCS) depressurization by opening power operated relief valves has important effect on hydrogen generation. Special attention should be paid to hydrogen generation enhancement caused by RCS depressurization.

关键词： pwr 核技术研究发展 RCS

Utilization of spent pwr fuel-advanced nuclear fuel cycle of pwr/CANDU synergism

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Nuclear Science and Techniques 2004年第3期15卷 183-187页

作者： HUOXiao-Dong XIEZhong-Sheng DepartmentofNuclearEngineering Xi＇anJiaotongUniversity.Xi＇an710049

High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both pwr and CANDU reactors and the closed cycle policy of reprocessing the spent pwr fuel is adopted, one of the advanced nuclear fuel cycles of pwr/CANDU synergism using the reprocessed uranium of spent pwr fuel in CANDU reactor is proposed, which will save the uranium resource (～22.5%), increase the energy output (～41%), decrease the quantity of spent fuels to be disposed (～2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent pwr fuel in CANDU reactor is a feasible and desirable strategy in China.

关键词：核燃料循环 pwr 乏燃料铀循环 CANDU

一种基于温控机制的pwr水冷包层TBR优化方法

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核聚变与等离子体物理 2017年第4期37卷 475-481页

作者：邱阳姚达毛张杰刘常乐高翔中国科学院等离子体物理研究所合肥230031 中国科学技术大学合肥230027

根据DEMO情形下的堆芯和包层设计参数,基于包层产氚及热工的温度控制要求,研究了一种适合pwr水冷包层设计的氚增殖率(TBR)优化方法。利用中子学及有限元方法,建立了包层2维分析模型,计算了包层能量增益因子,并基于该因子计算了包层的总... 详细信息

根据DEMO情形下的堆芯和包层设计参数,基于包层产氚及热工的温度控制要求,研究了一种适合pwr水冷包层设计的氚增殖率(TBR)优化方法。利用中子学及有限元方法,建立了包层2维分析模型,计算了包层能量增益因子,并基于该因子计算了包层的总能量沉积。研究了包层结构参数变化时局部TBR的分布特点。重点基于温控方法研究了包层结构参数的变化与TBR之间的内在联系,获得了包层局部TBR达到最佳值的结构参数理论参考值。

关键词： TBR优化温控机制 pwr 水冷包层 DEMO

Proposal of a Deuterium-Deuterium Fusion/pwr Fission Hybrid Reactor

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World Journal of Nuclear Science and Technology 2024年第4期14卷 190-233页

作者： Patrick Lindecker Independent Researcher Paris France

This article proposes to associate a Deuterium-Deuterium (D-D) fusion reactor with a pwr (fission Pressurized Water Reactor) in a hybrid reactor. Even if the mechanical gain (Q factor) of the D-D fusion reactor is below the unity and consequently consumes more energy than it supplies, due to the high energy amplification factor of the pwr fission reactor, the global yield is widely superior to 1. As the energy supplied by the fusion reactor is relatively low and as the neutrons supplied are mainly issued from D-D fusions (at 2.45 MeV), the problems of heat flux and neutrons damage connected with materials, as with D-T fusion reactors are reduced. Of course, there is no need to produce Tritium with this D-D fusion reactor. This type of reactor is able to incinerate any mixture of natural Uranium, natural Thorium and depleted Uranium (waste issued from enrichment plants), with natural Thorium being the best choice. No enriched fuel is needed. So, this type of reactor could constitute a source of energy for several thousands of years because it is about 90 more efficient than a standard fission reactor, such as a pwr or a Candu one, by extracting almost completely the energy from the fertile materials U238 and Th232. For the fission part, pwr technology is mature. For the fusion part, it is based on a reasonable hypothesis done on present Stellarators projects. The working of this reactor is continuous, 24 hours a day. In this paper, it will be targeted a reactor able to provide net electric power of about 1400 MWe, as a big fission power plant.

关键词： Fusion Reactor Fission Reactor Hybrid Reactor Nuclear Energy Deuterium-Deuterium Reactor Deuterium Colliding Beams Racetrack Stellarator Power Plant pwr

关于pwr及CANDU堆先进燃料管理策略的研究

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核动力工程 2000年第1期21卷 56-62页

作者：谢仲生西安交通大学 710049

阐述开展先进燃料管理策略研究的必要性与紧迫性。对我国秦山核电厂的燃料管理策略的改进进行了初步探讨 ,包括提高富集度延长循环长度、增大平均卸料燃耗、应用先进可燃毒物和低泄漏优化换料、改进燃料组件设计和适当提高功率等 ,并对... 详细信息

阐述开展先进燃料管理策略研究的必要性与紧迫性。对我国秦山核电厂的燃料管理策略的改进进行了初步探讨 ,包括提高富集度延长循环长度、增大平均卸料燃耗、应用先进可燃毒物和低泄漏优化换料、改进燃料组件设计和适当提高功率等 ,并对可能取得的重大经济效益进行了讨论。提出研究pwr的乏燃料在CNADU堆中应用及形成pwr/CANDU联合燃料循环的可行性 ,以提高燃耗深度 ,增加能量输出 ,降低发电成本。

关键词： pwr 燃料管理核电厂 CANDU堆燃料循环

RELAP5 Code Study of ROSA/LSTF Experiments on pwr Safety System Using Steam Generator Secondary-Side Depressurization

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Journal of Energy and Power Engineering 2015年第5期9卷 426-442页

作者： Takeshi Takeda Akira Ohnuki Hiroaki Nishi Japan Atomic Energy Agency Ibaraki-ken 319-1195 Japan Mitsubishi Heavy Industries Ltd. Hyogo-ken 652-8585 Japan The Kansai Electric Power Co. Inc. Fukui-ken 919-1141 Japan

RELAP5 （reactor excursion and leak analysis program, version 5） code analyses were performed on two ROSA/LSTF （rig of safety assessment/large scale test facility） experiments on pwr （pressurized water reactor） safety system that simulated cold leg small-break loss-of-coolant accidents with 8-in. or 4-in. diameter break using SG （steam generator） secondary-side depressurization. The SG depressurization was initiated by fully opening the depressurization valves in both SGs immediately after a safety injection signal. In the 8-in. break test, loop seal clearing occurred and then core uncovery and heatup took place by core boil-off. Core collapsed liquid level recovered after the initiation of accumulator coolant injection, and long-term core cooling was ensured by the actuation of low-pressure injection system. In the 4-in. break test, on the other hand, there was no core uncovery and heatup due to smaller break flow rate than in the 8-in. break test. Adjustment of Cd （break discharge coefficient） for two-phase discharge flow predicted the break flow rate reasonably well. The code well predicted the overall trend of the major thermal-hydraulic response observed in the two LSTF tests by the Cd adjustment. The code, however, overpredicted the peak cladding temperature because of underprediction of the core collapsed liquid level due to inadequate prediction of the accumulator flow rate in the 8-in. break case.

关键词： pwr safety system ROSAILSTF small-break loss-of-coolant accident SG depressurization RELAP5 code.

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pwr/CANDU联合核燃料循环研究

核科学与工程 2002年第3期22卷 256-260,277页

作者：谢仲生霍小东西安交通大学西安710049

根据我国已拥有pwr和CANDU核电站的具体情况 ,提出一种pwr/CANDU联合核燃料循环的策略 ,即把压水堆的乏燃料后处理后的回收铀 (RU)用作为CANDU堆的核燃料 ,既可节约铀资源 ,提高燃料的能量输出 ,又减少了废燃料的处置量 ,可大大降低核... 详细信息

根据我国已拥有pwr和CANDU核电站的具体情况 ,提出一种pwr/CANDU联合核燃料循环的策略 ,即把压水堆的乏燃料后处理后的回收铀 (RU)用作为CANDU堆的核燃料 ,既可节约铀资源 ,提高燃料的能量输出 ,又减少了废燃料的处置量 ,可大大降低核电成本。由于CANDU堆对核燃料循环的固有灵活性 ,堆芯结构及运行方式不需作重大改变 ,即可完成从天然铀到RU的过渡。又由于RU较低的放射性活度 ,这对CANDU堆的燃料制造是可以接受的 ,因而只需对现有燃料制造生产线稍加屏蔽措施 ,对运输和运行中燃料管理操作等都勿须改变。

关键词：联合核燃料循环 pwr CANDU 核电站核燃料管理

pwr核电站蒸汽发生器停堆湿保养工况联氨的缓蚀作用

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原子能科学技术 1992年第4期26卷 86-89页

作者：张孟琴潘庆春于晶华侯淑凤中国原子能科学研究院北京102413

一、引言压水堆(pwr)核电站蒸汽发生器(SG)管材因二回路系统腐蚀产物积累发生应力腐蚀开裂,这是SG传热管破损的主要原因之一。维修和更换SG使pwr停运期间所需要的替用电力对发电站造成很大的财政负担,同时,二回路系统的腐蚀产物沉积在... 详细信息

一、引言压水堆(pwr)核电站蒸汽发生器(SG)管材因二回路系统腐蚀产物积累发生应力腐蚀开裂,这是SG传热管破损的主要原因之一。维修和更换SG使pwr停运期间所需要的替用电力对发电站造成很大的财政负担,同时,二回路系统的腐蚀产物沉积在蒸汽发生器内。

关键词：联氨 pwr SG 停堆湿保养核电厂