Safety analysis of the reference research reactor MTR during reactivity insertion accident using the code MERSAT

The IAEA’s reference research reactor MTR-10 MW has been modeled using the code MERSAT. The developed MERSAT model consists of detailed representation of primary and secondary loops including reactor pool, bypass, main pump, heat exchanger and reactor core with the corresponding neutronics and thermalhydraulic characteristics. Following the successful accomplishment of the steady state operation at nominal power of 10 MW, reactivity insertion accident (RIA) for three different initial reactivity values of $1.5/0.5 s, $1.35/0.5 s and $0.1/1.0 s have been simulated. The predicted peaks of reactor power, hot channel fuel, clad and coolant temperatures demonstrate inherent safety features of the reference MTR reactor. Only in case of the fast RIA of $1.5/0.5 s, where the peak power of 133.66 MW arrived 0.625 s after the start of the transient, the maximum hot channel clad temperature arrives at the condition of subcooled boiling with the subsequent void formation. However, due to the strong negative reactivity feedback effects of coolant and fuel temperatures the void formation persists for a very short time so that thermalhydraulic conditions remained far from exceeding the safety design limits of thermalhydraulic instability and DNB. Finally, the simulation results show good agreement with previous international benchmark analyses accomplished with other qualified channel and thermalhydraulic system codes.     

CARR备用供电系统设计

备用供电系统是CARR安全运行的保障,它在两路外电源全都丧失时,自动启动两台柴油发电机组,向允许供电有短时中断的安全相关负荷提供电力。     

CARR上的应力测量中子衍射谱仪概念设计和模拟研究

中子衍射法是迄今为止可直接测量材料或工程部件内部深处应力场分布的唯一非破坏性方法，在工程上具有重要的应用。中国先进研究堆（CARR)中子散射工程拟建造一台应力测量中子衍射谱仪，其主要功能是测量材料中的残余应力和载荷应力。本文介绍了该谱仪的概念设计方案，并应用蒙特卡罗模拟软件MCSTAS对设计方案进行了模拟研究，对部分中子部件参数进行了优化设计。     

The evaluation of a reactivity insertion due to the group motion of FBR core subassemblies beyond Design Basis Ground Motion

The seismic probabilistic safety assessment (PSA) for fast breeder reactors (FBRs) has been carried out to confirm that the seismic safety is equivalent to that of light water reactors (LWRs). The seismic response on the reactor structure of FBRs causes seismic reactivity. The group motion of fuel assemblies is one of a typical seismic response. Therefore, much attention has been paid on the reactivity insertion mechanism due to the group motion of fuel assemblies and its consequence during the earthquake over the Design Basis Ground Motion (DBGM) condition. When the displacement of each subassembly is moving toward the same direction, each gap reduces coherently and the radial core compaction occurs, which results in positive reactivity insertion. We evaluate the gap reduction characteristics at the mid-plane of core by using a correlation coefficient. As a result, the fuel subassemblies are most concentrated when the input seismic motion of about 5 Hz frequency and 40 m/s² acceleration is applied. The amount of reactivity insertion is estimated approximately 1$ that corresponds to prompt criticality.     

聚变堆氚分析程序TAS1.0的设计与开发

氚是聚变堆的重要燃料之一,对聚变堆氚系统进行分析从而实行有效的氚控制是聚变研究的重要内容之一.在中国系列液态金属锂铅包层聚变堆概念设计研究基础上,利用现代软件工程方法及面向对象技术设计思想,发展了聚变堆氚分析程序TAS1.0,可用于聚变堆氚自持分析、氚燃料管理及氚安全性分析与研究,并可为聚变堆包层及燃料循环系统设计与分析提供技术支持.通过一系列的测试校验,表明了该程序的正确性与有效性.本文主要介绍该程序的系统设计、技术特点与程序测试.     

DINROS程序在中国实验快堆事故分析中的应用

DINROS是应用于多环路、多回路快中子反应堆装置瞬态工况分析计算的系统程序,也可以用于快中子反应堆动态特性及安全性能的研究.给出了DINROS程序在中国实验快堆事故分析中的应用.     

Development of DETAC and its application to the hydrogen detonation analysis

In this study, a numerical analysis code (DETAC, Detonation Analysis Code) for hydrogen detonation during the reactor severe accident was developed using Fortran 90 language, and the simulation was performed for the hydrogen detonation. A global-chemistry model was adopted to simulate the chemical reaction. The Euler equations were solved using third-order Runge-Kutta method with fifth-order weighted essentially non-oscillatory scheme handling the convection flux. Afterward, the hydrodynamics solver was verified by comparison of predicted results and exact solutions of four cases of shock tube problems. A hydrogen detonation in a pipe was simulated to verify this code by comparing the results with the classical C-J theory. Furthermore, this code was applied to the hydrogen detonation analysis in the compartment of BWR building. Two cases with different ignition locations were analyzed in this paper and the maximum pressure of these cases were 7.5 MPa and 8.0 MPa, respectively. The pressure and the temperature during detonation were affected by the ignition location. The results indicated that the possibility of reactor building destruction exists if the hydrogen detonation occurs.     

Improvement in estimation of first peak power based on non-linear temperature feedback reactivity in criticality accident with instantaneous reactivity insertion

A simple equation for the first peak power in a criticality accident due to instantaneous reactivity insertion into nuclear fuel solution system has been developed to improve the accuracy in the estimation of the first peak power keeping the easiness of calculation.

The equation is based on the assumption that temperature feedback reactivity is a second-order function of an increase in fuel temperature. Peak power estimated using the equation was in a range between about a half and twice of experimental value. Its applicability to a wide range of initial reactivity and accuracy of estimation have been confirmed in the comparison to one-point kinetics numerical calculation.

The expression suggests the first peak power increases with the square of small initial reactivity and three-halves power of large initial reactivity.     

中国先进研究堆安全设计

中国先进研究堆(CARR)是一座轻水冷却和慢化、重水反射的池内簟式研究堆。额定核功率为60MW。堆芯装载21盒燃料组件,芯体材料为U_3Si_2-Al_x弥散体,包壳材料为6061铝。CARR具有堆芯小、热流密度高和流速高等特点,使得CARR的安全设计难度很大。本文详细介绍了CARK设计中采取的安全措施,如ATWS缓解系统、足够大的主泵转动惯量、足够的自然循环能力和靠UPS供电的随堆运行的应急堆芯冷却系统等。事故分析结果表明,CARR具有很高的固有安全性,采取的安全措施是有效的。     

Research activities on nuclear reactor physics and thermal-hydraulics in Japan after Fukushima-Daiichi accident

Research and development in nuclear reactor physics and thermal-hydraulics continue to be vital parts of nuclear science and technology in Japan. The Fukushima accident not only brought tremendous change in public attitudes towards nuclear engineering and technology, but also had huge influence towards the research and development culture of scientific communities in Japan. After the Fukushima accident, thorough accident reviews were completed by independent committees, namely, Tokyo Electric Power Company (TEPCO), the Japanese government, the Diet of Japan, the Rebuild Japan Initiative Foundation, and the Nuclear and Industrial Safety Agency. Reactor physics and thermal-hydraulics divisions of Atomic Energy Society of Japan (AESJ) also issued the roadmaps after the accident. As a result, lessons learned from the accident were made clear, and a number of new research activities were initiated. The present paper reviews ongoing nuclear engineering research activities in Japanese institutes, universities, and corporations, focusing on the areas in reactor physics and thermal-hydraulics since the Fukushima accident to the present date.     

稳态堆芯多物理耦合系统CSSS V1.0的研发

充分考虑反应堆堆芯中子学物理、热工水力、燃料等专业的相互耦合过程,将先进节块法堆芯中子学计算软件NACK V1.0、热工水力子通道软件CORTH V2.0、燃料棒性能分析软件FUPAC V1.1进行集成耦合,得到稳态堆芯多物理耦合模拟设计分析系统CSSS V1.0,可计算典型压水堆的稳态运行物理、热工、燃料等专业参数。通过NEACRP-L-335压水堆基准问题验证计算,CSSS V1.0系统的计算结果与国际基准PARCS程序总体符合较好。     

Comprehensive and consistent interpretation of local fault experiments and application to hypothetical local overpower accident in Monju

Experimental studies on local fault (LF) accidents in fast breeder reactors have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Comprehensive and consistent interpretations of in-pile and out-of-pile experiments related to LF were arrived at in this study based on state-of-the-art review and data analysis techniques. Safety margins for a hypothetical local overpower accident, which was evaluated as a LF accident in the licensing document of the construction permit for a prototype fast breeder reactor called Monju, were also studied. Based on comprehensive interpretations of the latest experimental database, including those performed after the permission of Monju construction, it was clarified that the evaluation of the hypothetical local overpower accident in the Monju licensing was sufficiently conservative. Furthermore, it incorporated adequate safety margins in terms of failure thresholds of the fuel pin, molten fuel ejection, fuel sweep-out behavior after molten fuel ejection, and pin-to-pin failure propagation. Moreover, these comprehensive interpretations are valid and applicable to the safety evaluation of LF accidents of other fast breeder reactors with various fuel and core designs.     

核事故下放射性核素多介质耦合弥散模型

建立了高斯烟团模型与欧拉输运扩散模型的耦合模型,以实现放射性核素在大气以及水体中弥散的耦合。通过对大气和水体弥散模型进行对比,验证了各个模块的正确性。基于该模型,对中国铅基研究实验堆发生燃料组件熔化事故释放的~(131)I进行仿真分析,对该核素在大气以及水体中的浓度分布进行评估。模拟结果表明:事故发生2 h后大气污染主要分布在10 km内,水域污染主要分布在10~20 km处;在该模拟条件下,铅基堆发生燃料组件熔化事故后对大气和水体造成的影响均低于国家限值。     

严重事故条件下压力容器完整性评价的研究进展

堆芯熔融物堆内滞留(In-Vessel Retention,IVR)是以AP1000为代表的第三代轻水反应堆严重事故管理的重要策略之一,也是严重事故条件下保证压力容器完整性(Reactor Vessel Integrity,RVI)的典型方法之一.该文综述了国外在严重事故条件下压力容器完整性试验研究和理论分析的现状,总...     

中国先进研究堆(CARR)工程燃料组件设计验证试验

中国先进研究堆(CARR)采用的燃料组件在国内尚属首次加工与使用。为了保证燃料组件的完整性和安全性,满足堆安全运行的需要,对燃料板和组件的结构稳定性、流致振动、临界流速、热循环、堆内辐照等进行了设计验证试验。结果表明,CARR燃料组件的设计和加工工艺是合理的,谈组件在反应堆实际运行条件下是稳定和安全的。     

Study of safety features and accident scenarios in a fusion DEMO reactor

After the Fukushima Dai-ichi nuclear accident, a need for assuring safety of fusion energy has grown in the Japanese (JA) fusion research community. DEMO safety research has been launched as a part of Broader Approach DEMO Design Activities (BA-DDA). This paper reports progress in the fusion DEMO safety research conducted under BA-DDA. Safety requirements and evaluation guidelines have been, first of all, established based on those established in the Japanese ITER site invitation activities. The radioactive source terms and energies that can mobilize such source terms have been assessed for a reference DEMO concept. This concept employs in-vessel components that are cooled by pressurized water and built of a low activation ferritic steel (F82H), contains solid pebble beds made of lithium-titanate (Li₂TiO₃) and beryllium–titanium (Be₁₂Ti) for tritium breeding and neutron multiplication, respectively. It is shown that unlike the energies expected in ITER, the enthalpy in the first wall/blanket cooling loops is large compared to the other energies expected in the reference DEMO concept. Reference accident event sequences in the reference DEMO in this study have been analyzed based on the Master Logic Diagram and Functional Failure Mode and Effect Analysis techniques. Accident events of particular concern in the DEMO have been selected based on the event sequence analysis and the hazard assessment.     

OASIS程序的开发与应用

全面描述了对来自法国原子能委员会 (简称CEA)的快堆系统安全分析程序OASIS的引进和开发工作 ,并在此基础上介绍了该程序在中国实验快堆 (ChinaExperimentalFastReactor,简称CEFR)初步安全分析报告中对主给水管道断裂事故的分析计算。     

Development of safety assessment methodology on fuel element failure propagation in SFR and its application to Monju

Probabilistic and deterministic safety assessments and experimental studies on local fault (LF) propagation in sodium-cooled fast reactors (SFRs) have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Adventitious fuel pin failures have been considered to be the most dominant initiators of LFs in these probabilistic assessments because of its high frequency of occurrence during reactor operation and possibility of subsequent pin-to-pin failure propagation. Four possible mechanisms of fuel element failure propagation from adventitious fuel pin failure (FEFPA) were identified from a state-of-the-art review of open papers. All the mechanisms for FEFPA analysis including thermal, mechanical and chemical propagation are modeled into a safety assessment code which is applicable to arbitrary SFRs by developing some needed but missing methods. Furthermore, an assessment on FEFPA of Japanese prototype fast breeder reactor (Monju) was performed using this methodology. It was clarified that FEFPA was highly unlikely and limited at most within one subassembly in Monju owing to its redundant and diverse detection and shutdown systems for FEFPA even assuming the propagation. These results also suggested future possibility of run-beyond-cladding-breach operation which would enhance the economic efficiency in Monju.     

Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium-cooled fast reactors

The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt-through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior. The analysis method of molten material discharge was developed based on the computer code SIMMER-III since this code is designed to simulate the multi-phase, multi-component fluid dynamics with phase changes involved in the discharge process. Several experiments simulating the molten material discharge through duct using simulant materials were utilized as the basis of validation study of the physical models in this code. It was shown that SIMMER-III with improved physical models could simulate the molten material discharge behavior, including the momentum exchange with duct wall and thermal interaction with coolant. In order to develop an evaluation methodology of molten material penetration into sodium pool, a series of experiments simulating jet penetration behavior into sodium pool in SFR thermal condition were performed. These experiments revealed that the molten jet was fragmented in significantly shorter penetration length than the prediction by existing correlation for light water reactor conditions, due to the direct contact and thermal interaction of molten materials with coolant. The fragmented core materials form a sediment debris bed in the lower plenum. It is necessary to remove decay heat safely from this debris bed to achieve IVR. A simulation code to analyze the behavior of debris bed with decay heat was developed based on SIMMER-III code by implementing physical models, which simulate the interaction among solid particles in the bed. The code was validated by several experiments on the fluidization of particle bed by two-phase flow. These evaluation methodologies will serve as a basis for advanced safety assessment technology of SFRs in the future.     

Development of safety injection flow map associated with target depressurization for effective severe accident management of OPR1000

If any severe accident occurs, application of the Severe Accident Management Guidance (SAMG) is initiated by the Technical Support Center (TSC). In order to provide advisory information to the TSC, required safety injection flow rate for maintaining the coolability of the reactor core has been suggested in terms of the depressurization pressure. In this study, mechanistic development of the safety injection flow map was performed by post-processing the core exit temperature (CET) data from MELCOR simulation. In addition, effect of oxidation during the core degradation was incorporated by including simulation data of core water level decrease rate. Using the CET increase rate and core water level decrease rate, safety injection flow maps required for removing the decay and oxidation heat and finally for maintaining the coolability of the reactor core were developed. Three initiating events of small break loss of coolant accidents without safety injection, station black out, and total loss of feed water were considered. Reactor coolant system depressurization pressure targeting the suggested injection flow achievable with one or two high pressure safety injections was included in the map. This study contributes on improving the current SAMG by providing more practical and mechanistic information to manage the severe accidents.