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An old issue and a new challenge for nuclear reactor safety

F. D’AURIA

《能源前沿（英文）》 2021年第15卷第4期页码 854-859 doi: 10.1007/s11708-021-0729-0

摘要： Nuclear reactor safety (NRS) and the branch accident analysis (AA) constitute proven technologies: these are based on, among the other things, long lasting research and operational experience in the area of water cooled nuclear reactors (WCNR). Large break loss of coolant accident (LBLOCA) has been, so far, the orienting scenario within AA and a basis for the design of reactors. An incomplete vision for those technologies during the last few years is as follows: Progress in fundamentals was stagnant, namely in those countries where the WCNR were designed. Weaknesses became evident, noticeably in relation to nuclear fuel under high burn-up. Best estimate plus uncertainty (BEPU) techniques were perfected and available for application. Electronic and informatics systems were in extensive use and their impact in case of accident becomes more and more un-checked (however, quite irrelevant in case of LBLOCA). The time delay between technological discoveries and applications was becoming longer. The present paper deals with the LBLOCA that is inserted into the above context. Key conclusion is that regulations need suitable modification, rather than lowering the importance and the role of LBLOCA. Moreover, strengths of emergency core cooling system (ECCS) and containment need a tight link.

关键词： large break loss of coolant accident (LBLOCA) nuclear reactor safety (NRS) licensing perspectives basis for design of water cooled nuclear reactors (WCNR)

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Studies on advanced water-cooled reactors beyond generation III for power generation

CHENG Xu

《能源前沿（英文）》 2007年第1卷第2期页码 141-149 doi: 10.1007/s11708-007-0018-6

摘要： China s ambitious nuclear power program motivates the country s nuclear community to develop advanced reactor concepts beyond generation III to ensure a long-term, stable, and sustainable development of nuclear power. The paper discusses some main criteria for the selection of future water-cooled reactors by considering the specific Chinese situation. Based on the suggested selection criteria, two new types of water-cooled reactors are recommended for future Chinese nuclear power generation. The high conversion pressurized water reactor utilizes the present PWR technology to a large extent. With a conversion ratio of about 0.95, the fuel utilization is increased about 5 times. This significantly improves the sustainability of fuel resources. The supercritical water-cooled reactor has favorable features in economics, sustainability and technology availability. It is a logical extension of the generation III PWR technology in China. The status of international R&D work is reviewed. A new supercritical water-cooled reactor (SCWR) core structure (the mixed reactor core) and a new fuel assembly design (two-rows FA) are proposed. The preliminary analysis using a coupled neutron-physics/thermal-hydraulics method is carried out. It shows good feasibility for the new design proposal.

关键词： Chinese situation selection generation water-cooled feasibility

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Experience gained in analyzing severe accidents for WWER RP using CC SOCRAT

《能源前沿（英文）》 2021年第15卷第4期页码 872-886 doi: 10.1007/s11708-021-0796-2

摘要： The current Russian regulatory documents on the safety of nuclear power plant (NPP) specify the requirements regarding design basis accidents (DBAs) and beyond design basis accidents (BDBAs), including severe accidents (SAs) with core meltdown, in NPP design (NP-001-15, NP-082-07, and others). For a rigorous calculational justification of BDBAs and SAs, it is necessary to develop an integral CC that will be in line with the requirements of regulatory documents on verification and certification (RD-03-33-2008, RD-03-34-2000) and will allow for determining the amount of data required to provide information within the scope stipulated by the requirements for the structure of the safety analysis report (SAR) (NP-006-16). The system of codes for realistic analysis of severe accidents (SOCRAT) (formerly, thermohydraulics (RATEG)/coupled physical and chemical processes (SVECHA)/behavior of core materials relocated into the reactor lower plenum (HEFEST)) was developed in Russia to analyze a wide range of SAs at NPP with water-cooled water-moderated power-generating reactor (WWER) at all stages of the accident. Enhancements to the code and broadening of its applicability are continually being pursued by the code developers (Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN)) with OKB Gidropress JSC and other organizations. Currently, the SOCRAT/1 code can be used as a base tool to obtain realistic estimates for all parameters important for computational justification of the reactor plant (RP) safety at the in-vessel stage of SAs with fuel melting. To perform analyses using CC SOCRAT/1, the experience gained during execution of thermohydraulic codes is applied, which allows for minimizing the uncertainties in the results at the early stage of an accident scenario. This study presents the results of the work performed in 2010–2020 in OKB Gidropress JSC using the CC SOCRAT/1. Approaches have been considered to develop calculational models and analyze SAs using CC SOCRAT. This process, which is clearly structured in OKB Gidropress JSC, provides a noticeable reduction in human involvement, and reduces the probability of erroneous results.

关键词： system of codes for realistic analysis of severe accidents (SOCRAT) design basis accidents (DBAs) severe accidents (SAs) computer code (CC) nuclear power plant (NPP) design water-cooled water-moderated (WWER) modeling model safety requirements

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Dynamic simulation of a space gas-cooled reactor power system with a closed Brayton cycle

《能源前沿（英文）》 2021年第15卷第4期页码 916-929 doi: 10.1007/s11708-021-0757-9

摘要： Space nuclear reactor power (SNRP) using a gas-cooled reactor (GCR) and a closed Brayton cycle (CBC) is the ideal choice for future high-power space missions. To investigate the safety characteristics and develop the control strategies for gas-cooled SNRP, transient models for GCR, energy conversion unit, pipes, heat exchangers, pump and heat pipe radiator are established and a system analysis code is developed in this paper. Then, analyses of several operation conditions are performed using this code. In full-power steady-state operation, the core hot spot of 1293 K occurs near the upper part of the core. If 0.4 $ reactivity is introduced into the core, the maximum temperature that the fuel can reach is 2059 K, which is 914 K lower than the fuel melting point. The system finally has the ability to achieve a new steady-state with a higher reactor power. When the GCR is shut down in an emergency, the residual heat of the reactor can be removed through the conduction of the core and radiation heat transfer. The results indicate that the designed GCR is inherently safe owing to its negative reactivity feedback and passive decay heat removal. This paper may provide valuable references for safety design and analysis of the gas-cooled SNRP coupled with CBC.

关键词： gas-cooled space nuclear reactor power closed Brayton cycle system startup and shutdown positive reactivity insertion accident

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海洋核动力装备国内外发展现况与前景展望

郑洁,余凡,朱军民,柳存根,王欣月,朱英富

《中国工程科学》 2023年第25卷第3期页码 62-73 doi: 10.15302/J-SSCAE-2023.03.007

摘要：

海洋核动力装备是解决深远海资源开发中持久动力能源供给、海洋领域“碳减排”等问题的重要支撑。我国作为核电大国、海洋大国，虽然在核工业和海洋装备产业领域具有较好的优势基础，但在民用海洋核动力装备领域尚未实现“从零到一”的突破。本文基于对国内外海洋核动力装备发展实践研究，总结了海洋核动力装备的优势特性和技术策源，分析了未来海洋核动力装备发展的应用场景和主要趋势，厘清了我国发展海洋核动力装备的战略需求与问题，并提出了相关发展建议。研究认为海洋核动力装备总体呈现由军用向民用拓展、由陆地向海洋拓展的发展趋势，技术策源以紧凑型和一体化压水堆为主，装备类型近期将聚焦于海上浮动核电站和核动力破冰船。研究建议，通过顶层规划明确我国海洋核动力装备发展的重点应用场景，通过建立示范工程形成与发展需求相匹配的法规标准和监管制度等措施，突破海洋堆系统建造和核动力平台总装建造等方面的关键技术，推动海洋核动力装备高质量发展。

关键词：海洋核动力装备；小型模块化反应堆；压水堆；核动力船舶；海上浮动核电站

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Corrosion mechanisms of candidate structural materials for supercritical water-cooled reactor

Lefu ZHANG, Fawen ZHU, Rui TANG

《能源前沿（英文）》 2009年第3卷第2期页码 233-240 doi: 10.1007/s11708-009-0024-y

摘要： Nickel-based alloys, austenitic stainless steel, ferritic/martensitic heat-resistant steels, and oxide dispersion strengthened steel are presently considered to be the candidate structural or fuel-cladding materials for supercritical water-cooled reactor (SCWR), one of the promising generation IV reactor for large-scale electric power production. However, corrosion and stress corrosion cracking of these candidate alloys still remain to be a major problem in the selection of nuclear fuel cladding and other structural materials, such as water rod. Survey of literature and experimental results reveal that the general corrosion mechanism of those candidate materials exhibits quite complicated mechanism in high-temperature and high-pressure supercritical water. Formation of a stable protective oxide film is the key to the best corrosion-resistant alloys. This paper focuses on the mechanism of corrosion oxide film breakdown for SCWR candidate materials.

关键词： supercritical water-cooled reactor general corrosion oxide film corrosion mechanism

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中国新一代核能用材总体发展战略研究

干勇,赵宪庚,徐匡迪

《中国工程科学》 2019年第21卷第1期页码 1-5 doi: 10.15302/J-SSCAE-2019.01.001

摘要：

材料技术是支撑和保障核工程安全稳定运行的前提和基础。我国现有在役和在建的56台核电机组中有52台是压水堆，钠冷快堆和高温气冷堆正在开展示范工程电站的建设，其他堆型尚处于研究阶段。本文分析了我国新一代核能用材研发、制造、应用过程中存在的共性问题、在役和在建核能工程用材存在的突出问题、在研核能技术用材存在的关键问题，在此基础上提出了我国新一代核能用材的发展战略建议，包括设立国家新一代核能用材专业指导委员会，设立新一代核能用材国家专项基金或长期稳定支持的专项科技计划，创建我国新一代核能用材先进完整标准体系，建设国家层面的共享型工程级辐照实验装置，在独立自主原则下，继续开展新一代核能用材国际合作等。

关键词：新一代核能压水堆核能用材发展战略

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中国高温气冷堆制氢发展战略研究

张平,徐景明,石磊,张作义

《中国工程科学》 2019年第21卷第1期页码 20-28 doi: 10.15302/J-SSCAE-2019.01.004

摘要：

核能制氢是一种有应用前景的高效、大规模、无排放的制氢技术，有望在氢气大规模集中供应的场景中起到重要作用。高温气冷堆是最适于核能制氢的堆型，在我国已有几十年的研发基础，目前正在国家科技重大专项支持下建造高温气冷堆示范电站。本文介绍了核能制氢技术的特点和主流的核能制氢工艺包括热化学碘硫循环、混合硫循环和高温蒸汽电解的原理，对国际上核能制氢技术发展现状进行了简要综述，并概述了清华大学在该领域的研发现状。此外对核能制氢的安全性、技术经济评价等问题进行了讨论，在此基础上对与高温气冷堆耦合的制氢技术进行了评价，并以氢气直接还原炼铁为例探讨了高温气冷堆制氢在工业领域的应用前景。最后对我国高温气冷堆制氢技术的发展路线进行了探讨。

关键词：高温气冷堆能制氢热化学循环高温电解技术路线

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Preliminary design of an SCO conversion system applied to the sodium cooled fast reactor

《能源前沿（英文）》 2021年第15卷第4期页码 832-841 doi: 10.1007/s11708-021-0777-5

摘要： The supercritical carbon dioxide (SCO₂) Brayton cycle has become an ideal power conversion system for sodium-cooled fast reactors (SFR) due to its high efficiency, compactness, and avoidance of sodium-water reaction. In this paper, the 1200 MW_e large pool SFR (CFR1200) is used as the heat source of the system, and the sodium circuit temperature and the heat load are the operating boundaries of the cycle system. The performance of different SCO₂ Brayton cycle systems and changes in key equipment performance are compared. The study indicates that the inter-stage cooling and recompression cycle has the best match with the heat source characte-ristics of the SFR, and the cycle efficiency is the highest (40.7%). Then, based on the developed system transient analysis program (FR-Sdaso), a pool-type SFR power plant system analysis model based on the inter-stage cooling and recompression cycle is established. In addition, the matching between the inter-stage cooling recompression cycle and the SFR during the load cycle of the power plant is studied. The analysis shows that when the nuclear island adopts the flow-advanced operation strategy and the carbon dioxide flowrate in the SCO₂ power conversion system is adjusted with the goal of maintaining the sodium-carbon dioxide heat exchanger sodium side outlet temperature unchanged, the inter-stage cooling recompression cycle can match the operation of the SFR very well.

关键词： sodium-cooled fast reactor (SFR) supercritical carbon dioxide (SCO₂) brayton cycle load cycle

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Core designing of a new type of TVS-2M FAs: neutronics and thermal-hydraulics design basis limits

Saeed GHAEMI, Farshad FAGHIHI

《能源前沿（英文）》 2021年第15卷第1期页码 256-278 doi: 10.1007/s11708-018-0583-x

摘要： One of the most important aims of this study is to improve the core of the current VVER reactors to achieve more burn-up (or more cycle length) and more intrinsic safety. It is an independent study on the Russian new proposed FAs, called TVS-2M, which would be applied for the future advanced VVERs. Some important aspects of neutronics as well as thermal hydraulics investigations (and analysis) of the new type of Fas are conducted, and results are compared with the standards PWR CDBL. The TVS-2M FA contains gadolinium-oxide which is mixed with UO (for different Gd densities and U-235 enrichments which are given herein), but the core does not contain BARs. The new type TVS-2M Fas are modeled by the SARCS software package to find the PMAXS format for three states of CZP and HZP as well as HFP, and then the whole core is simulated by the PARCS code to investigate transient conditions. In addition, the WIMS-D5 code is suggested for steady core modeling including TVS-2M FAs and/or TVS FAs. Many neutronics aspects such as the first cycle length (first cycle burn up in terms of MW d/kgU), the critical concentration of boric acid at the BOC as well as the cycle length, the axial, and radial power peaking factors, differential and integral worthy of the most reactive CPS-CRs, reactivity coefficients of the fuel, moderator, boric acid, and the under-moderation estimation of the core are conducted and benchmarked with the PWR CDBL. Specifically, the burn-up calculations indicate that the 45.6 d increase of the first cycle length (which corresponds to 1.18 MW d/kgU increase of burn-up) is the best improving aim of the new FA type called TVS-2M. Moreover, thermal-hydraulics core design criteria such as MDNBR (based on W3 correlation) and the maximum of fuel and clad temperatures (radially and axially), are investigated, and discussed based on the CDBL.

关键词： TVS-2M FAs core design basis limits VVER-1000 analysis mixture of uranium-gadolinium oxides fuels thermal-hydraulics PARCS WIMS-D5

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Feasibility analysis of modified AL-6XN steel for structure component application in supercritical water-cooled

Xinggang LI, Qingzhi YAN, Rong MA, Haoqiang WANG, Changchun GE

《能源前沿（英文）》 2009年第3卷第2期页码 193-197 doi: 10.1007/s11708-009-0030-0

摘要： Modified AL-6XN austenite steel was patterned after AL-6XN superaustenitic stainless steel by introducing microalloy elements such as zirconium and titanium in order to adapt to recrystallizing thermo-mechanical treatment and further improve crevice corrosion resistance. Modified AL-6XN exhibited comparable tensile strength, and superior plasticity and impact toughness to commercial AL-6XN steel. The effects of aging behavior on corrosion resistance and impact toughness were measured to evaluate the qualification of modified AL-6XN steel as an in-core component and cladding material in a supercritical water-cooled reactor. Attention should be paid to degradation in corrosion resistance and impact toughness after aging for 50 hours when modified AL-6XN steel is considered as one of the candidate materials for in-core components and cladding tubes in supercritical water-cooled reactors.

关键词： supercritical water cooled reactor tensile impact toughness corrosion aging

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小型核反应堆是核电领域的下一个大事件吗？

Mitch Leslie

《工程（英文）》 2020年第6卷第3期页码 210-212 doi: 10.1016/j.eng.2020.01.003

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福岛核事故后中国广东核电集团核电厂抗震设计和评估进展

毛庆,吴应喜,张健,孟阿军,张涛,杨春菊,刘芳

《中国工程科学》 2013年第15卷第4期页码 46-51

摘要：

本文介绍了核电站抗震设计要求、在建和运行核电站的抗震设计情况以及运行核电站遭遇地震灾害的情况，简述了福岛核事故后世界各国核电站在抗震方面采取的措施，针对中国广东核电集团在福岛核事故后的行动进行了详细介绍，并提出了新建核电厂在抗震设计和评估方面的策略，以期通过技术手段持续提升核电站的抗震能力。

关键词：抗震设计基准超设计基准地震抗震裕量分析（SMA）隔震

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Achievement of high rate nitritation with aerobic granular sludge reactors enhanced by sludge recirculation

Zulkifly JEMAAT,Josep Anton TORA,Albert BARTROLI,Julián CARRERA,Julio PEREZ

《环境科学与工程前沿（英文）》 2015年第9卷第3期页码 528-533 doi: 10.1007/s11783-014-0641-5

摘要： A ratio control strategy has been used to demonstrate the feasibility of this automatic control procedure for the achievement of stable full and partial nitritation. The control strategy assured constant ratio between the dissolved oxygen (DO) and the total ammonia nitrogen (TAN) concentrations in the bulk liquid of aerobic granular sludge reactors operating in continuous mode. Three different set-ups with different reactor capacities were used (3, 110, and 150 L). High strength synthetic wastewaters and reject water were tested with similar performance. Achieved nitrogen loading rates ranged between 0.4 and 6.1 kgN·m ·d , at temperatures between 20°C and 30°C. Granular sludge and nitritation were stable in the long term continuous operation of the reactors. Suitable stable effluent for Anammox has been obtained using the desired TAN setpoint (i.e. 50% of influent ammonium oxidation). An existing biofilm model developed incorporating the implemented control loops and validated in a previous publication was used to investigate the effects of the ammonium concentration of the influent and the biofilm density on the achievement of full nitritation. The model demonstrated how sludge recirculation events led to a stable and significant increase of the biomass concentration in the reactor, which in turn resulted in the achievement of high nitrogen loading rates, due to the action of the control strategy. The model predicted an enhancement of stable full nitritation at higher ammonium concentrations in the influent. Poor influence of the biofilm density in the achievement of full nitritation was predicted with the model.

关键词： partial nitrification reject water high strength ammonium wastewater closed-loop control

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中国新一代核能核燃料总体发展战略研究

李冠兴,周邦新,肖岷,焦拥军,任忠鸣

《中国工程科学》 2019年第21卷第1期页码 6-11 doi: 10.15302/J-SSCAE-2019.01.002

摘要：

本文深入分析和研究了国内外压水堆燃料和材料技术，快堆及其他先进堆燃料技术以及核燃料循环相关材料技术发展的现状和趋势，提出了我国压水堆、快堆及其他先进堆核燃料与材料，以及核燃料循环材料发展的目标、发展路线图和重点任务。压水堆是我国21世纪相当长时间内核能发电及能源结构转型的主力堆型。作为压水堆发展重要支撑的核燃料及材料基本实现了国产化，但还没有实现品牌自主化。我国的快堆及快堆核燃料发展面临机遇和挑战，核燃料循环产业面临重大历史性发展机遇和巨大挑战。最后对我国的压水堆、快堆、其他先进堆型核燃料及材料，以及我国核燃料循环材料的发展提出了建议。

关键词：核燃料核材料轻水堆压水堆快堆燃料循环