民用小堆SGTR防满溢设计改进

民用小堆因单位功率下的蒸汽发生器（SG）汽空间偏小,稳压器容积和SG传热管内径偏大等特点,会引起蒸汽发生器传热管破裂（SGTR）事故快速满溢。本文采用RELAP5程序对民用小堆SGTR事故开展了优化措施研究,并提出极限单一故障下防止SG发生满溢的工程可行方案,即增加SG高水位排放液体的溢流管线或提高二次侧设计压力且同时增加自动的安注闭锁信号,保证在事故过程中蒸汽发生器不满溢和放射性排放满足限值要求。在民用小堆专设设备基本不变的前提下,针对系统进行了优化,极大地提升了安全性,为民用小堆设计改进提出了工程可行方案。     

小型堆稳压器波动管双端破口试验研究

开展了模块化小堆稳压器波动管双端破口试验研究,获得了非能动安全系统的事故响应特性和一回路系统参数变化。试验研究结果表明,在稳压器波动管双端破口极端工况条件下,中压安注箱能在短时间内提供较大的稳定安注流量,及时补充系统水装量;高压安注系统运行过程比较复杂,安注流量与堆芯补水箱压力平衡管线内介质状态和中压安注系统运行状态密切相关,在1.7 h内呈间歇注入运行状态。在整个事故过程中,堆芯一直处于淹没状态,模块化小堆非能动安全系统能够确保稳压器波动管在双端破口极端工况条件下的堆芯安全。      

核电厂仪控系统误动作事故分析

为全面评价核电厂仪控系统误动作事故,提出基于简化分析的方法,该方法基于功能组概念对仪控误动作假设始发事件（PIE）进行了系统化地识别和归并,得到不能被已有事故分析包络的潜在新增事故工况,并根据保守的分析假设和准则,针对识别出的潜在新增事故进行了定性评价和定量分析。研究结果表明,核电厂保护系统能够对仪控系统误动作事故提供多样化保护,事故后果满足验收准则,并建议增设“2个热管段实际压力与饱和压力之差低2信号触发安注启动以及所有主泵停运”功能。      

海洋工况下海洋小堆热工水力特性分析

海洋的开发,对核能有很大的需求空间.相较于陆基堆,海洋小堆在堆形的结构设计、设备及系统布置等,有着不同的特征.为适应船舶舱室与海洋环境,本文参照某型模块化小堆的主要设备结构和系统布置,得到一种改进后的海洋一体化小堆设计方案和设计参数.使用添加了海洋工况计算模块的RELAP5软件,对其中一个设计方案,引入海洋工况;通过和...     

模块式小堆全厂断电事故应对策略研究

研究了ACP100模块式小堆的全厂断电（SBO）事故应对策略,分析了非能动余热排出系统、非能动堆芯冷却系统对于SBO事故的缓解作用。研究结果表明,ACP100具有多种不依赖于可靠电源应对SBO事故的策略,ACP100采用非能动余热排出系统或非能动堆芯冷却系统均可以保障SBO事故下的堆芯余热长期导出,长期维持反应堆的可冷却性。      

模块式小堆超压风险及设计优化研究

模块式小堆采用带直流蒸汽发生器（OTSG）的一体化堆芯设计。OTSG具有传热面积大、设备体积小、蒸汽品质高的优点,然而因其二次侧水装量小、热惯性差,当反应堆发生二次侧排热减少时,反应堆冷却剂系统（RCS）可能存在超压风险。紧凑的一体化布置使得堆芯应对冷却剂受热膨胀的能力减弱,进一步增大RCS超压风险。本文采用RELAP5程序对模块式小堆的超压风险进行了研究。研究结果表明,模块式小堆在二次侧排热减少事故中会出现RCS超压现象,其中汽轮机事故停机导致的超压后果最为严重。波动管的流通面积对于RCS压力有着显著影响,合理地设计波动管流通面积可缓解RCS超压。     

应急撤离模拟在核电厂道路方案优化中的应用北大核心CSCD

将概率风险评价方法应用于核电厂的应急撤离模拟,利用自主编写的简化撤离模拟程序,结合厂址事故源项、人口、道路、气象条件等特征,对多个核电厂应急撤离条件下公众与工作人员可能的受照剂量和风险进行了对比分析。在此基础上,结合霞浦厂址应急道路方案遇到的实际问题,在保证事故应急状态下公众和工作人员能够有效撤离的同时,对应急道路方案进行了比选,为工程的实施提供借鉴和参考。相关程序和方法也可为后续发展海岛核电、小型供热堆等提供技术支持,有助于更直观地开展核电公众沟通。     

小堆发展的法规标准问题分析及建议

当前小堆从设计研发阶段正逐步推向市场应用。小堆自身特点和设计理念与传统大型核动力堆不尽一致,这对我国当前的法规标准的优化和改进提出了挑战。文章介绍了法规标准在小堆方面的安全要求,分析了当前小堆发展的法规标准存在的共性问题,并提出了我国法规标准支持小堆发展方面的优化和改进建议,可供我国核电行业法规标准后续建设进行参考。     

反应堆一回路系统优化设计方案的可行性验证

采用优选运行参数和结构参数的方法,可达到降低核动力装置尺寸的目的。在优化设计方案投入制造前,有必要研究其在设计基准事故下的响应特性,以检验优化方案的可行性。采用RELAP5/MOD3.2程序研究现有一回路系统优化方案在完全失去厂外电、主给水丧失和小破口失水事故下的响应特性,并将安全设计准则参数与母型对比。结果表明：针对所研究的3种设计基准事故,优化方案各主要安全准则参数满足设计要求;优化方案可成功抵御这3类设计基准事故。     

实验快堆自然循环模拟实验装置的模拟原理及其参数计算

研究了池式快堆自然循环模拟实验的模拟准则，根据模拟准则和自然循环守衡方程式，对池式实验快堆自然循环模拟实验装置，在各种模拟准则条件下的几何与热工设计参数进行了计算。研究了模型比例，事故冷却器一次侧进出口温差和阻力系数等对相似准则数的影响，并且确定了模拟实验装置的设计参数范围，从理论上解决了池式实验快堆自然循环模拟实验装置的模拟问题。     

使用分级方案应用研究堆应急准备与响应要求研究

我国当前的研究堆应急管理没有对不同类别研究堆的应急准备与响应作出差异性要求,分级方案是根据与反应堆相关的潜在危害正当应用这些安全要求的良好手段。按照分级方案的步骤,基于我国当前研究堆安全分类准则、国际原子能机构（IAEA）功率相关应急威胁分类准则以及应用IAEA应急准备与响应要求的分级方案的依据,提出了我国研究堆的应急管理分类准则以及对不同应急管理类别研究堆应急状态分级和应急计划区（EPZ）要求,这为简化低功率研究堆营运单位应急预案的内容和细节的范围、程度和水平以及建立与不同类别研究堆危害评定结果相称的我国研究堆应急管理系统提供了依据。      

Design of the VVR-S research reactor

A water-cooled, water-moderated reactor for facilitating scientific research endeavors on applications of nuclear energy in peaceful pursuits has been built in the Soviet Union.Such reactors are currently completed and in operation in the Soviet Union and in other Socialist countries. Six such reactors were put into operation during 1957–1959; five reactors (four of which are built to handle power surges) are in the stage of preparation, assembly, and start-up tests.This article describes the design of the VVR-S reactor and its experimental facilities. The physical characteristics of the reactor have been described in an earlier paper [1].     

核电厂烟羽应急计划区划分方法研究

结合不同类型反应堆的安全特性,对不同的烟羽应急计划区(PEPZ)划分方法进行对比分析,然后依据反应堆类型进行系统归类,提出不同堆型所适用的烟羽应急计划区划分方法,最后以我国的模块式高温气冷堆示范电厂(HTR-PM)为例进行划分方法的初步应用。初步研究结果表明,HTR-PM在厂址边界处满足烟羽应急计划区的划分准则,相对于目前的大型轻水堆,可以明显减小其烟羽应急计划区。     

Preliminary study on feasibility of large and small water cooled thorium breeder reactor in equilibrium states

Fuel breeding is one of the essential performances for a self-sustaining reactor system which can maintains the fuel sustainability while the reactor produces energy and consumes the fissile materials during operation. Thorium cycle shows some advantageous on higher breeding characteristics in thermal neutron spectrum region as shown in the Shippingport reactor and molten salt breeder reactor (MSBR) project. In the present study, the feasibility of large and small water cooled thorium breeder reactors is investigated under equilibrium conditions where the reactors are fueled by ²³³U–Th oxide and they adopts light water coolant as moderator. The key properties such as required enrichment, breeding capability, and initial fissile inventory are evaluated. The conversion ratio and fissile inventory ratio (FIR) are used for evaluating breeding performance. The results show the feasibility of breeding for small and large reactors. The breeding performance increases with increasing power output and lower power density. The small reactor may achieve the breeding condition when the fuel pellets' power density of about 22.5 W/cm³ and burnup of about 20 GWd/t.     

陆上小型堆烟羽应急计划区划分初步研究

小型堆烟羽应急计划区(EPZ)大小作为其市场推广和应用的重要外部约束条件之一,意味着制定合适的划分准则和确立其大小范围具有十分重大的意义。结合现行大堆烟羽应急计划区(EPZ)的划分准则,本文分析了国内外小型堆烟羽应急计划区(EPZ)划分方法,提出陆上小型堆采用剂量/距离的划分方法。在研究中,基于MAAP程序对某小型堆进行建模计算,从中得出了较为合理的机理性应急源项;并通过大气扩散计算软件MACCS程序进行烟羽应急计划区(EPZ)计算;同时对厂址差异进行相关的灵敏性分析。     

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。     

Thermohydraulic studies of sodium-cooled reactor facilities

Thermohydraulic studies of reactor facilities with fast reactors are complex experimentally and computationally. Extensive experimental data are obtained on the velocity and temperature profiles, hydrodynamic resistance and heat emission, initial heat section, and interchannel mixing of the coolant in the fuel assemblies. These are used to develop engineering methods of performing thermohydraulic calculations of fuel assemblies as well as computational compute codes. The particulars of the hydrodynamics and heat transfer in intermediate heat exchangers and steam generators of reactor facilities with fast reactors are studied. This has made it possible to validate their thermohydraulic characteristics.     

Fuel storage facilities for 200-MW nuclear heating reactor

This paper reflects the thoughts and work concerning considerations and design of 200-MW nuclear heating reactor (NHR-200) developed in Institute of Nuclear Energy Technology (INET), Tsinghua University, China. Due to the fact that the size of heating reactors is limited to the local demands which are generally smaller that the economic reasonable size as compared to those reactors for electricity production, the design of systems for NHR-200 should be specified in accordance with its design characteristics, and simplified as much as possible for economic aim. The nuclear heating reactor has a low power density in the core and that the annual generation period is only about 180 days. Therefore, the total required number of fuel bundles is rather small. Furthermore, in-vessel spent fuel storage is feasible. All these features raise the potential to simplify the fuel storage system for NHR-200. The fuel storage and inspection facilities are described.     

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.     

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.