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Influence evaluation of loading conditions during pressurized thermal shock transients based on thermal-hydraulics

Jinya KATSUYAMA, Shumpei UNO, Tadashi WATANABE, Yinsheng LI

《机械工程前沿（英文）》 2018年第13卷第4期页码 563-570 doi: 10.1007/s11465-018-0487-9

摘要：

The thermal hydraulic (TH) behavior of coo-lant water is a key factor in the structural integrity assessments on reactor pressure vessels (RPVs) of pressurized water reactors (PWRs) under pressurized thermal shock (PTS) events, because the TH behavior may affect the loading conditions in the assessment. From the viewpoint of TH behavior, configuration of plant equipment and their dimensions, and operator action time considerably influence various parameters, such as the temperature and flow rate of coolant water and inner pressure. In this study, to investigate the influence of the operator action time on TH behavior during a PTS event, we developed an analysis model for a typical Japanese PWR plant, including the RPV and the main components of both primary and secondary systems, and performed TH analyses by using a system analysis code called RELAP5. We applied two different operator action times based on the Japanese and the United States (US) rules: Operators may act after 10 min (Japanese rules) and 30 min (the US rules) after the occurrence of PTS events. Based on the results of TH analysis with different operator action times, we also performed structural analyses for evaluating thermal-stress distributions in the RPV during PTS events as loading conditions in the structural integrity assessment. From the analysis results, it was clarified that differences in operator action times significantly affect TH behavior and loading conditions, as the Japanese rule may lead to lower stresses than that under the US rule because an earlier operator action caused lower pressure in the RPV.

关键词： structural integrity reactor pressure vessel pressurized thermal shock thermal hydraulic analysis pressurized water reactor weld residual stress

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中国改进型压水堆核电技术——CPR1000的形成

濮继龙

《中国工程科学》 2008年第10卷第3期页码 54-57

摘要：

介绍了目前正在建设的岭澳核电站（二期）和辽宁红沿河核电站所采用技术——CPR1000的形成过程及其中的重要技术创新点。

关键词：压水堆 CPR1000 自主品牌

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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

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压水堆熔融物堆内滞留策略：历史回顾与研究展望 Review

马卫民,元一单,Bal Raj Sehgal

《工程（英文）》 2016年第2卷第1期页码 103-111 doi: 10.1016/J.ENG.2016.01.019

摘要：

本文对广泛应用于第三代压水堆的严重事故缓解措施——熔融物堆内滞留(IVR)进行了历史回顾。IVR策略最早源自于第二代反应堆Lovissa VVER-440的改进设计，以应对堆芯熔化事故。随后，IVR策略被应用于许多新设计的反应堆，如西屋的AP1000、韩国的APR1400以及中国的先进压水堆CAP1400和华龙一号。对IVR策略有效性影响最大的因素分别为堆内堆芯熔化进程、熔融物加载于压力容器壁面的热流密度和压力容器外部冷却。对于堆芯熔化进程，过去人们一直仅关注压力容器下腔室内熔池的换热行为。但通过回顾与分析，本文认为堆内的其他现象，如堆芯的降级和迁移、碎片床的形成及其可冷却性以及熔池的动态形成过程等，可能也会对熔池的最终状态及其作用于下封头的热负荷产生影响。通过对相关研究的回顾，本文希望找出IVR策略的研究中有待完善的部分，并据目前发展水平提出未来IVR研究的需求。

关键词：压水堆严重事故熔融物堆内滞留碎片床形成碎片床再熔化熔池形成熔池热工水力学临界热流密度

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HPR1000：具备能动与非能动安全性的先进压水堆 Review

邢继,宋代勇,吴宇翔

《工程（英文）》 2016年第2卷第1期页码 79-87 doi: 10.1016/J.ENG.2016.01.017

摘要：

HPR1000是具有能动与非能动安全性的先进核电站。它是基于现有压水堆核电站成熟技术的渐进式设计，融合了包括采用177组CF3先进燃料组件的堆芯能动与非能动安全系统、全面的严重事故预防与缓解措施、强化的外部事件防护能力和改进的应急响应能力在内的先进设计特征。针对关键的自主创新技术，如非能动系统、堆芯和主设备，研究人员已经开展了充分的试验验证。 HPR1000的设计满足国际上对先进轻水堆的用户要求以及最新的核安全要求，并且考虑了福岛事故的经验反馈。基于其出色的安全性与经济性，HPR1000为国内与国际核电市场提供了卓越可行的解决方案。

关键词： HPR1000 能动与非能动安全性先进核动力堆(先进压水堆)

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CFD Simulation of thermal hydraulic characteristics in a typical upper plenum of RPV

Mingjun WANG, Lianfa WANG, Yingjie WANG, Wenxi TIAN, Jian DENG, Guanghui SU, Suizheng QIU

《能源前沿（英文）》 2021年第15卷第4期页码 930-945 doi: 10.1007/s11708-021-0728-1

摘要： A comparative computational fluid dynamics (CFD) study was conducted on the three different types of pressurized water reactor (PWR) upper plenum, named TYPE 1 (support columns (SCs) and control rod guide tubes (CRGTs) with two large windows), TYPE 2 (SCs and CRGTs without windows), and TYPE 3 (two parallel perforated barrel shells and CRGTs). First, three types of upper plenum geometry information were collected, simplified, and adopted into the BORA facility, which is a 1/5 scale system of the four-loop PWR reactor. Then, the geometry, including the upper half core, upper plenum region, and hot legs, was built using the platform. After that, an unsteady calculation to simulate the reactor balance operation at hot full power scenario was performed. Finally, the differences of flowrate distribution at the core outlet and temperature distribution and transverse velocity inside the hot legs with different upper plenum internals were compared. The results suggest that TYPE 1 upper plenum internals cause the largest flowrate difference at the core outlet while TYPE 3 leads to the most even distributed flowrate. The distribution and evolution pattern of the tangential velocity inside hot legs is highly dependent on the upper plenum internals. Two counter-rotating swirls exist inside the TYPE 1 hot leg and only one swirl revolving around the hog leg axis exist inside the TYPE 2 hot leg. For TYPE 3, two swirls like that of TYPE 1 rotating around the hot leg axis significantly increase the temperature homogenization speed. This research provides meaningful guidelines for the future optimization and design of advanced PWR upper plenum internal structures.

关键词： pressurized water reactor (PWR) upper plenum internal structures temperature distribution computational fluid dynamics (CFD)

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常压核供热——技术现实经济可行的清洁能源

田嘉夫

《中国工程科学》 2000年第2卷第2期页码 74-76

摘要：

为治理城市的燃煤污染和降低二氧化碳的排放量，我国需要开发大型清洁能源。核能是可供选择的清洁能源之一，核能不仅可以用于发电，而且可以替代燃煤为城市供热。在我国现在池式研究堆技术的基础上，设计出的常压供热反应堆，可以满足城市供热的需要，具有高度的安全性和可靠性，是一种技术现实和经济可行的解决方案。在我国发展城市集中供热的条件下，采用这种大型热源，将会以较大的规模来改善大气环境和降低温室气体的排放。

关键词：大气污染温室气体核能供热

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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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特低渗透油田提压注水效果探讨与实践

张洪亮

《中国工程科学》 2012年第14卷第4期页码 62-64

摘要：

大庆西部外围地区特低渗透扶余油层具有埋藏深、厚度薄、砂体零散的特点，针对首次注水开发的特低渗透油田注水井吸水能力差、难以有效开发动用的实际，在搞清该区块微观特征的基础上，为确保特低渗透储层达到建立有效驱替，“注够水、注好水”的目的，经过反复理论研究和论证，提出了单井增压注水试验，现场应用后注水井能够完成配注，周围油井明显见到增油效果，实现了特低渗透储层有效驱替。利用无源微地震和同位素监测资料分析，改善了吸水状况，缓解了层间矛盾，储层动用程度得到提高。随注水压力升高降低，注水波及区范围随之升高、降低，增压注

关键词：增压注水无源微地震特低渗透油田波及系数有效驱替

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Experimental study of critical flow of water at supercritical pressure

Yuzhou CHEN, Chunsheng YANG, Shuming ZHANG, Minfu ZHAO, Kaiwen DU, Xu CHENG

《能源前沿（英文）》 2009年第3卷第2期页码 175-180 doi: 10.1007/s11708-009-0029-6

摘要： Experimental studies of the critical flow of water were conducted under steady-state conditions with a nozzle 1.41 mm in diameter and 4.35 mm in length, covering the inlet pressure range of 22.1-26.8 MPa and inlet temperature range of 38-474°C. The parametric trend of the flow rate was investigated, and the experimental data were compared with the predictions of the homogeneous equilibrium model, the Bernoulli correlation, and the models used in the reactor safety analysis code RELAP5/MOD3.3. It is concluded that in the near or beyond pseudo-critical region, thermal-dynamic equilibrium is dominant, and at a lower temperature, choking does not occur. The onset of the choking condition is not predicted reasonably by the RELAP5 code.

关键词： critical flow supercritical water-cooled reactor(SCWR) reactor safety loss of coolant accident(LOCA)

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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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Degradation of endocrine disruptor bisphenol A in drinking water by ozone oxidation

XU Bin, GAO Naiyun, RUI Min, WANG Hong, WU Haihui

《环境科学与工程前沿（英文）》 2007年第1卷第3期页码 350-356 doi: 10.1007/s11783-007-0060-y

摘要： The ozone oxidation of endocrine disruptor bisphenol A in drinking water was investigated. A stainless completely mixed reactor was employed to carry out the degradation experiments by means of a batch model. With an initial concentration of 11.0 mg/L, the removal efficiencies of BPA (bisphenol A) could be measured up to 70%, 82%, and 90% when the dosages of ozone were 1, 1.5, and 2 mg/L, respectively. The impacts on BPA degradation under the conditions of different ozone dosages, water background values, BPA initial concentrations, and ozone adding time were analyzed. The results showed that ozone dosage plays a dominant role during the process of BPA degradation, while the impact of the contact time could be ignored. UV wavelength scanning was used to confirm that the by-products were produced, which could be absorbed at UV254. The value of UV254 was observed to have changed during the ozonation process. Based on the change of UV254, it could be concluded that BPA is not completely degraded at low ozone dosage, while shorter adding time of total ozone dosage, high ozone dosage, and improvement of dissolved ozone concentration greatly contribute to the extent of BPA degradation. The effects of applied HO dose in ozone oxidation of BPA were also examined in this study. The O-HO processes proved to have similar effects on the degradation of BPA by ozone oxidation.

关键词： BPA scanning reactor process O-HO

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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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Interlayer-confined two-dimensional manganese oxide-carbon nanotube catalytic ozonation membrane for efficient water

《化学科学与工程前沿（英文）》 2022年第16卷第5期页码 731-744 doi: 10.1007/s11705-021-2110-6

摘要： Catalytic ozonation technology has attracted copious attention in water purification owing to its favorable oxidative degradation of pollutants and mitigation of membrane fouling capacity. However, its extensive industrial application has been restricted by the low ozone utilization and limited mass transfer of the short-lived radical species. Interlayer space-confined catalysis has been theoretically proven to be a viable strategy for achieving high catalytic efficiency. Here, a two-dimensional MnO₂-incorporated ceramic membrane with tunable interspacing, which was obtained via the intercalation of a carbon nanotube, was designed as a catalytic ozonation membrane reactor for degrading methylene blue. Benefiting from the abundant catalytic active sites on the surface of two-dimensional MnO₂ as well as the ultralow mass transfer resistance of fluids due to the nanolayer confinement, an excellent mineralization effect, i.e., 1.2 mg O_3(aq) mg^–1 TOC removal (a total organic carbon removal rate of 71.5%), was achieved within a hydraulic retention time of 0.045 s of pollutant degradation. Further, the effects of hydraulic retention time and interlayer spacing on methylene blue removal were investigated. Moreover, the mechanism of the catalytic ozonation employing catalytic ozonation membrane was proposed based on the contribution of the Mn(III/IV) redox pair to electron transfer to generate the reactive oxygen species. This innovative two-dimensional confinement catalytic ozonation membrane could act as a nanoreactor and separator to efficiently oxidize organic pollutants and enhance the control of membrane fouling during water purification.

关键词： catalytic membrane reactor catalytic ozonation nanoconfinement two-dimensional manganese oxide

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Intrinsic kinetics and external diffusion of catalytic steam gasification of fine coal char particles under pressurized

Xuantao Wu, Jie Wang

《化学科学与工程前沿（英文）》 2019年第13卷第2期页码 415-426 doi: 10.1007/s11705-018-1725-8

摘要： Catalytic steam gasification of fine coal char particles was carried out using a self-made laboratory reactor to determine the intrinsic kinetics and external diffusion under varying pressures (0.1–0.5 MPa) and superficial gas flow velocities (GFVs) of 13.8–68.8 cm?s . In order to estimate the gas release rate at a low GFV, the transported effect of effluent gas on the temporal gasification rate pattern was simulated by the Fluent computation and verified experimentally. The external mass transfer coefficients ( ) and the effectiveness factors were determined at lower GFVs, based on the intrinsic gasification rate obtained at a high GFV of 55.0 cm?s . The was found to be almost invariable in a wider carbon conversion of 0.2–0.7. The variations of at a median carbon conversion with GFV, temperature and pressure were found to follow a modified Chilton-Colburn correlation: (0.04< <0.19), where is total pressure and is atmospheric pressure. An intrinsic kinetics/external diffusion integrating model could well describe the gasification rate as a function of GFV, temperature and pressure over a whole gasification process.

关键词： coal char catalytic steam gasification pressure kinetics diffusion