Air ingress phenomena in a depressurization accident of the very-high-temperature reactor

The inherent properties of the very-high-temperature reactor (VHTR) facilitate the design of the VHTR with high degree of passive safe performances, compared to other type of reactors. However, it is still not clear if the VHTR can maintain a passively safe function during the primary-pipe rupture accident, or what would be a design criterion to guarantee the VHTR with the high degree of passively safe performances during the accident. The primary-pipe rupture accident is one of the most common of accidents related to the basic design regarding the VHTR, which has a potential to cause the destruction of the reactor core by oxidizing in-core graphite structures and to release fission products by oxidizing graphite fuel elements. It is a guillotine type rupture of the double coaxial pipe at the nozzle part connecting to the side or bottom of the reactor pressure vessel, which is a peculiar accident for the VHTR. If a primary pipe ruptures, air will be entered into the reactor if there is air in the reactor containment or confinement vessels. This study is to investigate the air ingress phenomena and to develop the passively safe technology for the prevention of air ingress and of graphite corrosion. The present paper describes the influences of a localized natural circulation in parallel channels onto the air ingress process during the primary-pipe rupture accident of the VHTR.     

美国核管会关于小型模块堆厂址人口要求的考虑及启示

本文介绍了美国核管理委员会(NRC)关于反应堆选址过程中与人口因素相关的审管要求和评估准则,详细分析了NRC针对小型模块堆对涉及厂址人口要求法规的4种修订方案,提出我国在制定小型堆厂址人口要求过程中需要关注的问题:(1)考虑到总体社会风险并从厂址比选的角度出发,建立一个恰当的反应堆距人口集中居住区(或人口中心)边界的距离是必要的。此外,从纵深防御考虑,小型堆厂址仍然需要与人口集中居住区保持一个适当的距离;(2)基于小型堆选址事故后果及影响范围,建立与大型商用核动力厂相同社会风险水平的评价指标(如事故工况下厂址周围的集体有效剂量)是有益的;(3)公众可接受性和选址政策也是小型堆能否靠近人口集中居住区的重要因素。     

Update of the French R&D strategy on gas-cooled reactors

Gas-cooled reactors take up a strong second role in France's R&D strategy on future nuclear energy systems as priority was given in 2005 to fast neutron reactors with multiple-recycle for their potential to optimally use uranium resource and minimize the long term burden of radioactive waste. Owing to the European past experience on sodium-cooled fast reactors (SFRs), this reactor type was logically selected as reference for a new generation fast neutron reactor intended to be tested as a prototype in the 2020s and be ready for industrial deployment around 2040. At the same time, the potential merits of a gas fast reactor (GFR) with ceramic clad fuel for a safe management of cooling accident are acknowledged for the potential of this reactor type to resolve critical issues of liquid fast reactors (safety, operability and reparability). A pre-feasibility report on a first concept of GFR was issued in 2007 that summed-up results of a 5-year international R&D effort on GFR fuel technology, reactor design and operating transient analyses. This report established a global confidence in the feasibility of this concept and its potential for attractive performances. Furthermore, it suggested directions of R&D to generate by 2012 an updated concept with improved performances and taking better benefit from GFR specific technologies.A second activity on gas-cooled reactors originates from the current interest of CEA's industrial partner AREVA in high or very high temperature reactors (V/HTR) for supplying hydrogen, synthetic hydrocarbon fuels and process heat for the industry. This activity currently encompasses R&D on V/HTR key technologies such as particle fuel fabrication, high temperature compact heat exchangers and coupling technologies to various power conversion systems. R&D on V/HTR and GFR are synergistic in various respects. The GFR can be viewed as a more sustainable version of the VHTR and synergies exist in research on heat resisting materials, helium system technology and power conversion systems. Both reactors require active research in materials and spur developments of new metallic alloys and ceramics applicable to other advanced nuclear systems.     

Specifics of RBMK core cooling in beyond design basis accidents

The most dangerous beyond design basis accidents for RBMK reactors, leading to the worst consequences, are related to the loss of long-term heat removal from the core. Due to a specific design of RBMK, there are a few possibilities for heat removal from reactor core by non-regular means: removal of heat from graphite stack by reactor gas circuit, removal of heat from reactor core using control rods cooling circuit, depressurisation of reactor cooling system, supply of water into cooling system from low pressure water sources, etc. This paper presents the analysis of such heat removal by employing RELAP5, RELAP5-3D and RELAP/SCDAPSIM codes. The analysis was performed for Ignalina nuclear power plant with RBMK-1500 reactor. The analysis of result shows that the restoration of water supply into control rod channels enables to remove 10-30 MW of the generated heat from the reactor core. This amount of removed heat is comparable with reactor decay heat in long-term period and allows to slowdown the core heat-up process. However, the injection of water to reactor cooling system is considered as main strategy, which should be considered in RBMK-1500 accident management procedure.     

Neutronic and safety aspects of inert matrix fuel utilization in fast reactors for plutonium and minor actinides transmutation

The solving of ecological problems of future nuclear power is connected with the solving of long-lived radioactive waste utilization problems. It concerns primarily plutonium and minor actinides (MAs), accumulated in the spent fuel of nuclear reactors. One of the ways this can be solved is to use a fast reactor with uranium-free or inert matrix fuel (IMF). The physics of this type of reactor was widely investigated during last year for BN-800 reactors. The solution of the most important problems was: a decrease in non-uniformity of power distribution and an increase of the Doppler effect. The next stage of such core investigations is an evaluation of self-protection to beyond design accidents. Preliminary results show a high safety level of BN-800 reactors with IMF in the event of unprotected loss of coolant flow (ULOF) accident.     

对滨河研究堆堆址设计基准洪水的探讨

通过对有关标准条文的理解以及对现有水文计算技术和方法、研究堆水灾事故危害程度的分析,结合工程设计实例,对滨河研究堆堆址设计基准洪水提出了明确的标准,即较高功率研究堆应该具有与核电厂相同的防洪基准.     

Comparative study on safety performance of nitride fueled lead-bismuth cooled fast reactor with various power levels

In the present study safety performance of nitride fueled lead-bismuth cooled fast reactors of several sizes (150 MWt 2500 MWt) but all having maximum burn-up of about 911 % HM are evaluated and compared. Small reactors can be operated up to 12 years, and large reactors(2500MWt) can be operated up to about 4 years without refueling or fuel shuffling. In each reactor excess reactivity is minimized up to below β_eff in order to eliminate super-prompt critical accident. The ULOF and UTOP accident simulation was performed for each design and the results showed that all reactors could survive both accidents passively/inherently. However the temperature margin, especially for cladding, is larger for smaller reactor.     

10 MW固态燃料熔盐实验堆安全分析关键技术初步研究

钍基熔盐反应堆(Thorium Molten Salt Reactor,TMSR)项目是中国科学院科技先导项目之一。基于10 MW热功率熔盐反应堆-固体燃料(Thorium Molten Salt Reactor-Solid Fuel,TMSR-SF)的设计,对TMSR的关键技术安全分析进行了初步研究。TMSR-SF与现有反应堆之间的差异对核安全审查提出挑战,TMSR-SF审查方法的研究将准备其安全审查的技术和要求。固态燃料熔盐实验堆安全分析关键技术初步研究包含4个方面:堆芯核设计关键安全限值、事故序列及验收准则、源项及其审评方法和验收准则、概率安全评价方法和始发事件。首先对其它类型反应堆的安全审查方法进行了研究,对其关键参数和重要规定做了概述,并借鉴了高温气体冷堆和钠冷却快堆的审评要求和方法;然后使用蒙特卡罗和其他方法、模型来计算TMSR-SF的关键参数。应用逻辑图方法讨论概率风险评价(Probabilistic Risk Assessment,PRA)方法和始发事件清单。在本研究中,计算了核心核设计安全限值,研究和讨论事故列表和分类,讨论了TMSR-SF的PRA框架和始发事件清单,该研究将支持TMSR-SF的安全审查和安全设计。     

Safety characterstics of the MARS nuclear plant

New design criteria in the field of small and medium sized nuclear reactors will be illustrated; particular consideration will be given to the reactor MARS (Multipurpose Advanced Reactor, Inherently Safe) now under development at the Department of Energetics, University of Rome.An analysis of plant behaviour during a Station Black-out accident has been performed using the computer program RELAP 5/Mod. 2.     

Safety evaluation on the depressurization accident in the gas turbine high temperature reactor (GTHTR300)

Japan Atomic Energy Agency has been developing a gas turbine high temperature reactor (GTHTR300) with electric power of approximately 300 MW. One of the unique safety design concepts of this system is that events with frequency of occurrence of higher than 10⁻⁸/reactor-year are evaluated as design basis events in order to show that the frequency of large amount of FP release is less than 10⁻⁸/reactor-year. According to this concept, a depressurization accident by a large break of helium piping is postulated as a design basis event. This accident is one of the most serious accidents in the high-temperature gas-cooled reactors from the viewpoint of loss of coolability. The safety evaluation on the accident was conducted based on the actual design of the system. The short-term and long-term behaviors of fuel temperature after occurrence of the accident, internal pressure of the reactor building, oxidation behavior of fuels and graphite structures were evaluated and exposure dose of general public was also estimated using the results of evaluation of fuel temperature and fuel failure by oxidation. All of the evaluation results meet the safety criteria and feasibility of the GTHTR300 was shown by this study.     

模块化小型核反应堆自动卸压系统分析研究

模块化小型核反应堆（SMR）与传统大型压水堆在结构上存在很大差异,导致两者的严重事故进程存在较大差异。因此本文结合SMR自身设计特点,建立反应堆严重事故分析模型,对SMR的典型事故瞬态进行模拟计算,并对严重事故进程、热工水力现象和系统安全进行研究。在此基础上提出了SMR自动卸压系统优化改进方案,通过对自动卸压系统各级卸压管线的位置和阀门有效面积进行深入研究,并对相关参数进行敏感性分析,提出符合反应堆自身特点的卸压阀门有效面积的优化设计方案,为小型核反应堆的严重事故预防和缓解提供有效的依据和参考。     

Risk-informed design guidance for future reactor systems

Future reactor designs face an uncertain regulatory environment. It is anticipated that there will be some level of probabilistic insights in the regulations and supporting regulatory documents for Generation-IV nuclear reactors. Central to current regulations are design basis accidents (DBAs) and the general design criteria (GDC), which were established before probabilistic risk assessments (PRAs) were developed. These regulations implement a structuralist approach to safety through traditional defense in depth and large safety margins. In a rationalist approach to safety, accident frequencies are quantified and protective measures are introduced to make these frequencies acceptably low. Both approaches have advantages and disadvantages and future reactor design and licensing processes will have to implement a hybrid approach. This paper presents an iterative four-step risk-informed methodology to guide the design of future-reactor systems using a gas-cooled fast reactor emergency core cooling system as an example. This methodology helps designers to analyze alternative designs under potential risk-informed regulations and to anticipate design justifications the regulator may require during the licensing process. The analysis demonstrated the importance of common-cause failures and the need for guidance on how to change the quantitative impact of these potential failures on the frequency of accident sequences as the design changes. Deliberation is an important part of the four-step methodology because it supplements the quantitative results by allowing the inclusion in the design choice of elements such as best design practices and ease of online maintenance, which usually cannot be quantified. The case study showed that, in some instances, the structuralist and the rationalist approaches were inconsistent. In particular, GDC 35 treats the double-ended break of the largest pipe in the reactor coolant system with concurrent loss of offsite power and a single failure in the most critical place as the DBA for the emergency core cooling system. Seventeen out of the 45 variations that we considered violated this DBA, but passed the probabilistic screening criteria. Using PRA techniques, we found that the mean frequency of this accident was very low, thus indicating that deterministic criteria such as GDC 35 must be reassessed in the light of risk insights.     

Late-phase melt conditions affecting the potential for in-vessel retention in high power reactors

If cooling is inadequate during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 accident. In such a case, concerns about containment failure and associated risks can be eliminated if it is possible to ensure that the lower head remains intact so that relocated core materials are retained within the vessel. Accordingly, in-vessel retention (IVR) of core melt as a key severe accident management strategy has been adopted by some operating nuclear power plants and planned for some advanced light water reactors. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements can provide sufficient heat removal to assure IVR for high power reactors (i.e., reactors with power levels up to 1500 MWe). Consequently, a joint United States/Korean International Nuclear Energy Research Initiative (I-NERI) has been launched to develop recommendations to improve the margin of success for in-vessel retention in high power reactors. This program is initially focussed on the Korean Advanced Power Reactor—1400 MWe (APR1400) design. However, recommendations will be developed that can be applied to a wide range of existing and advanced reactor designs. The recommendations will focus on modifications to enhance ERVC and modifications to enhance in-vessel debris coolability. In this paper, late-phase melt conditions affecting the potential for IVR of core melt in the APR1400 were established as a basis for developing the I-NERI recommendations. The selection of ‘bounding’ reactor accidents, simulation of those accidents using the SCDAP/RELAP5-3D^© code, and resulting late-phase melt conditions are presented. Results from this effort indicate that bounding late-phase melt conditions could include large melt masses (>120,000 kg) relocating at high temperatures (3400 K). Estimated lower head heat fluxes associated with this melt could exceed the maximum critical heat flux, indicating additional measures such as the use of a core catcher and/or modifications to enhance external reactor vessel cooling may be necessary to ensure in-vessel retention of core melt.     

Conceptual design of multi-purpose heat reactor “Nuclear Heat Generator”

Small heat reactors can apply to on site demand such as district heat and air conditioning, industrial process heat, greenhouse, and seawater desalination in urban and rural areas. The purpose of this paper is to design conceptually a multi-purpose reactor named “Nuclear Heat Generator (NHG)” which could be installed in energy consuming area. The reactor of 1MWt output is designed without any needs for fuel exchange and decommissioning on site. This cassette typed reactor vessel with sealing is transported to specified fuel fabrication shop every 3 to 4 years in order to exchange used fuels. Steam generators are involved in the self-pressurized integrated reactor with natural circulation. Generated steam pressure from heating reactor is 0.88 MPa (saturated) which is so less than that of current water reactors. Under low steam pressure it is considerably easy to make design of containment vessel and safety device. For economic competition overcoming scale demerit it will be necessary for the cassette type reactor to optimize its system design for the multi-production effect as well as modular construction and recycling system.     

Level 1 probabilistic safety assessment to support the design of the CEA 2400 MWth gas-cooled fast reactor

As part of the design of a 4th generation reactor, the integration of safety in the early phase of the concepts is expected. To date, probabilistic insights are increasingly employed in the safety demonstration in combination with the deterministic approach (e.g. to identify the sequences of complex failures, to justify the categorization of situations) and used, even at an early stage of design, to identify the reliability of systems and equipment to handle the safety objectives (expressed in terms of core damage frequency targets). Within this frame, the CEA has undertaken to assess the benefit of developing a probabilistic model to support the design of the 2400 MWth gas-cooled fast reactor.In the building process of this level 1 probabilistic safety assessment, a first phase consisted in making a preliminary model that took only into account families of initiating events that were defined for the design of the decay heat removal dedicated loops, namely the loss of coolant accidents (representative of medium pressure situations) and loss of off-site power/station black-out transients (representative of high pressure situations). Owing to the results obtained with this preliminary L1PSA model, it emerged that an increased reliability of the DHR function in high pressure conditions (i.e. characterized by IEs not associated to the loss of integrity of the helium pressure boundary) is suitable to reduce the overall core damage frequency. The track was therefore chosen to require the use of normal loops as first line of provision of the DHR function, possibly including components or particular operating modes related to the secondary and tertiary circuits. In addition, this final L1PSA model is characterized by success criteria based on transient calculations performed with the CATHARE2 code and to a perimeter extended to all representative internal IEs at full operating power.This paper presents the building process and the main results related to these two successive L1PSA models. Finally, useful insights are translated into GFR design improvements that are leading to an overall CDF at full operating power that satisfies nowadays with the probabilistic target defined for 3rd generation reactors, being at least the objective for 4th generation reactors.     

Cyclic mode of neptunium, americium and curium transmutation in heavy-water reactor

Characteristics of process of transmutation of neptunium, americium and curium from spent nuclear fuel in heavy-water reactor during first 10 lifetimes and at transition to equilibrium mode are calculated. During transmutation, dangerous nuclides, first of all, ²⁴⁴Cm and ²³⁸Pu are accumulated. They cause an increase of radiotoxicity. At first 10 cycles of transmutation, the radiotoxicity is increased by 8.7 times in comparison with radiotoxicity of initial load of transmuted actinides. Heavy-water reactor with thermal power of 1000 MW can transmute neptunium, americium and curium extracted from 3.7 VVER-1000 type reactors. It means, that the required power of transmutation reactor makes about 8% of thermal power of VVER-1000 type reactors.     

Modelica仿真技术在气冷式微型反应堆中的应用

相比传统大型核电厂,微型反应堆各系统功能间紧密耦合且相互制约,传统的分专业解耦设计模式难以应对,需开展全范围的系统仿真。采用Modelica语言建立了气冷式微型反应堆的系统仿真模型,以未能紧急停堆的预期瞬态（ATWS）事故为例开展事故分析计算,并与专业堆芯安全分析结果对比,结果表明反应堆功率变化趋势较为一致,且ATWS事故后仅依靠堆芯温度升高引入的负反应性可实现停堆。本文研究方法为气冷式微型反应堆的全系统建模仿真打下了坚实基础,也为其他类型反应堆的系统建模仿真提供了很好的借鉴作用。      

Technical design features and essential safety-related properties of the HTR-module

An HTR-Module power plant consists of standardized reactor units with a power rating of 200 MW each, so that the special, inherent safety features of small high-temperature reactors are also applicable to power plants of any desired power rating. The main design features of the reactor and vital components are given. Special emphasis is layed on a comprehensive survey of design basis accidents and the corresponding accident doses in the environment. It is shown that these doses are much lower than the maximum allowable values given in the German Radiation Protection Ordinance even if it is postulated that the installed filters of the reactor building are not available.Finally, a summary of essential properties and design principles of the HTR-Module is given. These properties and principles were fully approved and accepted by the advisory experts (TÜV) of the German Licensing Authorities in their final assessment report on the conceptual design of the HTR-Module (November 1989).     

Inherent Prevention and Mitigation of Severe Accident Consequences in Sodium-Cooled Fast Reactors

Safety challenges for sodium-cooled fast reactors include maintaining core temperatures within design limits and assuring the geometry and integrity of the reactor core. Due to the high power density in the reactor core, heat removal requirements encourage the use of high-heat-transfer coolants such as liquid sodium. The variation of power across the core requires ducted assemblies to control fuel and coolant temperatures, which are also used to constrain core geometry. In a fast reactor, the fuel is not in the most neutronically reactive configuration during normal operation. Accidents leading to fuel melting, fuel pin failure, and fuel relocation can result in positive reactivity, increasing power, and possibly resulting in severe accident consequences including recriticalities that could threaten reactor and containment integrity. Inherent safety concepts, including favorable reactivity feedback, natural circulation cooling, and design choices resulting in favorable dispersive characteristics for failed fuel, can be used to increase the level of safety to the point where it is highly unlikely, or perhaps even not credible, for such severe accident consequences to occur.     

全厂断电事故工况下小型铅铋快堆余热排出能力评价

小型铅铋快堆的非能动余热排出系统（PRHRS）主要是为应对全厂断电（SBO）事故,但目前并不确定该PRHRS能否有效带走堆芯衰变热以保证堆芯安全,因此开展了数值分析研究评价PRHRS的余热排出能力。本文使用RELAP5 4.0程序开展了小型铅铋快堆SBO事故热工水力分析,首先进行稳态计算,之后将稳态结果作为初值进行瞬态计算。研究结果表明：在整个SBO事故中,包壳峰值温度最高为820 K,主容器与保护容器壁面最高温度分别为792 K和769 K,均未超过安全限值,表明此PRHRS可有效应对小型铅铋快堆SBO事故。本文研究可为小型铅铋快堆PRHRS的工程设计奠定技术基础。