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

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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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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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 MWe 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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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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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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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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A thermoelectric generator and water-cooling assisted high conversion efficiency polycrystalline silicon

Zekun LIU, Shuang YUAN, Yi YUAN, Guojian LI, Qiang WANG

《能源前沿（英文）》 2021年 第15卷 第2期   页码 358-366 doi: 10.1007/s11708-020-0712-1

摘要： Solar energy has been increasing its share in the global energy structure. However, the thermal radiation brought by sunlight will attenuate the efficiency of solar cells. To reduce the temperature of the photovoltaic (PV) cell and improve the utilization efficiency of solar energy, a hybrid system composed of the PV cell, a thermoelectric generator (TEG), and a water-cooled plate (WCP) was manufactured. The WCP cannot only cool the PV cell, but also effectively generate additional electric energy with the TEG using the waste heat of the PV cell. The changes in the efficiency and power density of the hybrid system were obtained by real time monitoring. The thermal and electrical tests were performed at different irradiations and the same experiment temperature of 22°C. At a light intensity of 1000 W/m , the steady-state temperature of the PV cell decreases from 86.8°C to 54.1°C, and the overall efficiency increases from 15.6% to 21.1%. At a light intensity of 800 W/m , the steady-state temperature of the PV cell decreases from 70°C to 45.8°C, and the overall efficiency increases from 9.28% to 12.59%. At a light intensity of 400 W/m , the steady-state temperature of the PV cell decreases from 38.5°C to 31.5°C, and the overall efficiency is approximately 3.8%, basically remain unchanged.

关键词： photovoltaic (PV)     thermoelectric generator     conversion efficiency     hybrid energy systems     water-cooled plate (WCP)    

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我国高温气冷堆发展战略研究

张作义,吴宗鑫,王大中,童节娟

《中国工程科学》 2019年 第21卷 第1期   页码 12-19 doi: 10.15302/J-SSCAE-2019.01.003

摘要：

高温气冷堆和在此基础上发展起来的超高温气冷堆是第四代核能系统研发重点的6种堆型之一。本文介绍了高温气冷堆的特点，对高温气冷堆技术在国内外的最新研发进展进行了简要综述，对高温气冷堆的发展定位等问题进行了讨论。在此基础上对我国高温气冷堆发展路线进行了展望。我国高温气冷堆技术历经跟踪、跨越和自主创新，目前在商业规模模块式高温气冷堆核电站技术上处于世界领先地位。在此基础上，我国正在启动部署后续60万千瓦级模块式高温气冷堆核电机组的研发和配套关键技术的攻关工作，以进一步推动高温气冷堆技术的产业化，保持我国在该领域的国际领先优势。

关键词： 高温气冷堆     高温     技术路线    

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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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Heat transfer with water flowing upward in a tube for pressures up to supercritical region

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

《能源前沿（英文）》 2010年 第4卷 第3期   页码 358-365 doi: 10.1007/s11708-009-0071-4

摘要： A heat transfer experiment was conducted in a tube of 6.07mm in diameter with water flowing upward, covering the ranges of pressure of 10―23MPa, mass flux of 288―1298kg/(m·s), local water temperature of 78°C―270°C, heat flux of 0.23―1.18MW/m and Reynolds number of 5.5×10―3.9×10. The experimental results were compared with the predictions of the Dittus-Boelter correlation, Jackson correlation, Bishop correlation, Swenson correlation and Yamagata correlation. Significant deterioration in heat transfer was observed in both subcritical and supercritical region due to the effect of buoyancy force, but it was not predicted reasonably by the existing correlations.

关键词： heat transfer     deterioration     buoyancy     supercritical water    

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山东石岛湾200 MWe 球床模块式高温气冷堆(HTR-PM) 核电站示范工程 Review

张作义, 董玉杰, 李富, 张征明, 王海涛, 黄晓津, 李红, 刘兵, 吴莘馨, 王宏, 刁兴中, 张海泉, 王金华

《工程（英文）》 2016年 第2卷 第1期   页码 112-118 doi: 10.1016/J.ENG.2016.01.020

摘要：

世界首台球床模块式高温气冷堆(HTR-PM) 核电站示范工程于2012 年12 月9日在中国山东省荣成市石岛湾厂区完成第一罐混凝土的浇筑，2015年6月完成反应堆厂房建设，然后进入设备安装阶段。目前正在向着在2017年年底实现并网发电的目标顺利推进。1个HTR-PM反应堆模块的热功 率是250 MWth，反应堆堆芯氦气的进出口温度分别是250 °C 和750 °C。蒸汽发生器出口的蒸汽参数是13.25 MPa/567 °C。2个球床反应堆模块连接1台蒸汽轮机，形成一座210 MWe的核电站。项目团队克服了巨大困难，利用中国现有的工业制造技术研制出世界首台设备，实现了一系列重大技术创新。在研发的规划和实施、工业合作伙伴关系的建立、主设备制造、燃料生产、安全审查、站址选择以及安全性和经济性的平衡等方面取得了令人欣慰的进展，为世界同行积累了可以借鉴的经验。

关键词： 核能     高温气冷堆     球床     模块式高温气冷堆     球床模块式高温气冷堆    

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Chemical reactions of oily sludge catalyzed by iron oxide under supercritical water gasification condition

《化学科学与工程前沿（英文）》 2022年 第16卷 第6期   页码 886-896 doi: 10.1007/s11705-021-2125-z

摘要： Supercritical water gasification is a promising technology in dealing with the degradation of hazardous waste, such as oily sludge, accompanied by the production of fuel gases. To evaluate the mechanism of Fe2O3 catalyst and the migration pathways of heteroatoms and to investigate the systems during the process, reactive force field molecular dynamics simulations are adopted. In terms of the catalytic mechanisms of Fe2O3, the surface lattice oxygen is consumed by small carbon fragments to produce CO and CO2, improving the catalytic performance of the cluster due to more unsaturated coordination Fe sites exposed. Lattice oxygen combines with •H radicals to form water molecules, improving the catalytic performance. Furthermore, the pathway of asphaltene degradation was revealed at an atomic level, as well as products. Moreover, the adsorption of hydroxyl radical on the S atom caused breakage of the two C–S bonds in turn, forming •HSO intermediate, so that the organic S element was fixed into the inorganic liquid phase. The heteroatom O was removed under the effects of supercritical water. Heavy metal particles presented in the oily sludge, such as iron in association with Fe2O3 catalyst, helped accelerate the degradation of asphaltenes.

关键词： oily sludge     SCWG     ReaxFF     Fe2O3     heteroatoms    

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Development of MCBurn and its application in the analysis of SCWR physical characteristics

Ganglin YU , Kan WANG ,

《能源前沿（英文）》 2009年 第3卷 第3期   页码 348-352 doi: 10.1007/s11708-009-0031-z

摘要： The MCBurn, a coupled code system linking the Monte Carlo N-particle transport code(MCNP) and Oak Ridge isotope generation and depletion code (ORIGEN), can resolve the basic calculation problems in reactor physical design and analysis, such as criticality, power distribution, nuclide burn up, and neutron fluence. It has been verified in the pressurized water reactor (PWR) pin model, fast reactor (FR) burn up model, and boiling water reactor(BWR) assemble model with benchmarked results. In supercritical water reactor (SCWR) physical calculations, the calculation conditions such as the geometry, the neutron spectrum, and the fuel materials, etc., are more complex than those in traditional reactors, which is a great challenge to reactor physics calculation code. However, the MCBurn code is a possible solution. In this paper, several update functions of the MCBurn in new neutron cross-section lib, code interface, and neutron flux distribution were described. The application of the MCBurn in SCWR were verified on a supercritical water reactor assemble. The calculation results show that the MCBurn is accurate and adaptable in the SCWR calculation.

关键词： Monte Carlo method     MCBurn     SCWR     neutron cross-section lib    

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标题 作者 时间 类型 操作

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

Lefu ZHANG, Fawen ZHU, Rui TANG

期刊论文

Experimental study of critical flow of water at supercritical pressure

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

期刊论文

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

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

期刊论文

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

CHENG Xu

期刊论文

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

期刊论文

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

期刊论文

An old issue and a new challenge for nuclear reactor safety

F. D’AURIA

期刊论文

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

期刊论文

A thermoelectric generator and water-cooling assisted high conversion efficiency polycrystalline silicon

Zekun LIU, Shuang YUAN, Yi YUAN, Guojian LI, Qiang WANG

期刊论文

我国高温气冷堆发展战略研究

张作义,吴宗鑫,王大中,童节娟

期刊论文

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

期刊论文

Heat transfer with water flowing upward in a tube for pressures up to supercritical region

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

期刊论文

山东石岛湾200 MWe 球床模块式高温气冷堆(HTR-PM) 核电站示范工程

张作义, 董玉杰, 李富, 张征明, 王海涛, 黄晓津, 李红, 刘兵, 吴莘馨, 王宏, 刁兴中, 张海泉, 王金华

期刊论文

Chemical reactions of oily sludge catalyzed by iron oxide under supercritical water gasification condition

期刊论文

Development of MCBurn and its application in the analysis of SCWR physical characteristics

Ganglin YU , Kan WANG ,

期刊论文