西安脉冲堆核事故场外后果概率安全评价

为定量评价西安脉冲堆(XAPR)场外风险,建立XAPR核事故场外后果概率评价模型,以XAPR场区特征气象数据为输入数据,分析计算了XAPR核事故场外后果。结果表明:完整释放谱发生后,在XAPR场区100 m边界处有效剂量超过1、10 m Sv的条件概率分别约为0.652%、0.0750%;个人有效剂量超过10 m Sv的总频率小于2.20×10-9 a-1;致死癌症风险超过1×10-6的总频率小于1.89×10-6 a-1;XAPR场外个人平均癌症死亡风险满足草拟的核安全目标。XAPR场外风险极小。     

西安脉冲堆辐射剂量监测

为了确保反应堆运行人员和公众的安全，西安脉冲堆设置了比较完善的辐射剂量监测系统。该系统包括固定式在线连续监测和便携式仪表及取样监测两部分，本文主要了固定式在线连续监测的监测,测点布置以及各监测仪的性能特点，并简要介绍了便携式仪表及取样监测中各仪表的性能与适用范围。     

西安脉冲堆脉冲参数测量装置研制

介绍了我国自主研制的第一套商用脉冲反应堆脉冲参数测量装置（西安脉冲堆脉冲参数测量装置）的设计方案、系统组成、工作原理、技术特点和应用情况。该装置是西安脉冲反应堆主控室控制台的主要仪表之一，具有脉冲参数测量和堆保护功能。装置采用微电子和计算机技术，实现了对反应堆脉冲工况下堆功率变化、脉冲波形与多个参数的实时自动测量与显示，并在功率超限时向反应堆保护系统和报警系统发出信号。文中对装置的工作特性做了描述，并对现场调试和运行情况作了简要介绍。     

西安脉冲堆概率安全分析技术要点及分析框架研究

针对西安脉冲堆（XAPR）自身设计特点及安全特性,研究了XAPR概率安全分析（PSA）的技术特殊要点,提出了XAPR PSA分析框架及技术要素具体实施方法。最后以XAPR堆水池中破口失水事故为始发事件,验证了XAPR PSA研究思路。分析表明:以始发事件为起点、事件序列为主干、放射性释放类为终点的一体化事件树结构分析框架适合于XAPR PSA。      

西安脉冲堆的调试管理

介绍了西安脉冲堆调试管理的模式和特点。西安脉冲堆调试的日常活动由调试领导小组负责，对试验项目启动的控制主要采取“调试试验卡”制度。编制了调试所需技术性和管理性程序，明确了参加调试的各部门的职责，制定了调试活动的主要原则，试验过程中分别对试验时间、试验范围进行控制。     

西安脉冲堆控制棒控制监视系统

西安脉冲堆控制棒驱动机构采用步进电机作为执行部件，与其相适应的控制棒控制系统采用了单片机芯片电路和大功率VMOS管，控制棒位置探测器显示系统采用接近开关和数字显示。经综合试验和堆上运行考验证明，该系统设计合理、性能良好、精确度高，具有很好的推广价值。     

西安脉冲堆瞬态运行试验研究

介绍了西安脉冲堆脉冲和方波的瞬态运行试验方法及结果。对脉冲参数测量结果进行了简要分析，结果表明，本堆瞬态运行总体性能良好，达到了预期的设计目标。     

西安脉冲堆脉冲参数分析

介绍西安脉冲堆堆芯装载布置,发射脉冲的过程和机理,脉冲参数的计算模型和程序以及计算值和实验测量结果分析。当引入2.478×10-2反应性进行脉冲运行时,实测脉冲峰功率为4301.3MW,达到了设计指标。西安脉冲堆总体性能达到了国外同类堆型的先进水平。     

西安脉冲堆脉冲控制棒驱动机构

西安脉冲堆脉冲控制棒驱动机构是实现反应堆脉冲运行的关键设备。该机构采用滚珠丝杠副的传动方式，通过气压缸实现控制棒的脉冲发射。脉冲控制棒驱动机构具有全行程脉冲时间短、结构简单、维修方便的特点。试验表明：脉冲控制棒驱动机构最大负荷大于300N，全行程脉冲时间小于100ms，落棒时间小于1.2s，寿命大于4000次脉冲发射运行，平均无故障时间大于400次脉冲发现运行。该机构已成功应用于西安脉冲堆。     

西安脉冲堆仿真系统

介绍了西安脉冲堆仿真系统的组成、原理、功能及特点,并就系统软件及仿真模型没计作了初步说明。该仿真系统不仅模拟脉冲堆两种堆芯布置如4种运行方式的操作,而且模拟了氙毒、碘坑过程及脉冲堆可能的7种事件,也可由教员随机设置多种设备故障或异常工况。该仿真系统已成功地用于操纵员培训及核工程专业学生的实习。     

一种基于~(103)Rh(n,n′)~(103)Rh~m反应的快中子注量监测方法

给出了一种基于~(103)Rh(n,n′)~(103)Rh~m反应监测快中子注量的方法。根据干扰核素半衰期不同的特点,选取合适的冷却和测量时间,降低了~(103)Rh(n,n′)~(103)Rh~m活化率测量中干扰核素的影响。根据152Eu标定实验结果,利用遗传算法优化探测器尺寸,建立探测器蒙特卡罗算法模型。利用模型校正~(99)Mo-~(99)Tcm放射源实验效率,解决了探测器效率标定和射线自吸收问题。开展了快中子注量监测,实验结果与铁、硫、镍等快中子监测箔一致性较好,测量不确定度约为13.1%。     

船舶核动力装置安全目标的初步研究

安全目标是核电厂进行安全评价的重要基础和判定准则,对安全评价工作具有非常重要的影响.目前,核电厂安全目标的认识和制定已经经历了较长的时间,形成了以国际原子能机构(IAEA)和美国核管会(NRC)为主的两大体系.文章概要地介绍了两个组织所确定的定性和定量安全目标,以及我国核电厂安全目标的发展和应用现状.最后,在吸收上述经...     

Monte Carlo simulation of a research reactor with nominal power of 7 MW to design new control safety rods

The Monte Carlo simulation has been established for a research reactor with nominal power of 7 MW. A detailed model of the reactor core was employed including standard and control fuel elements, reflectors, irradiation channels, control rods, reactor pool and thermal column. The following physical parameters of reactor core were calculated for the present LEU core: core reactivity (ρ), control rod (CR) worth, thermal and epithermal neutron flux distributions, shutdown margin and delayed neutron fraction. Reduction of unfavorable effects of blockage probability of control safety rod (CSR)s in their interiors because of not enough space in their sites, and lack of suitable capabilities to fabricate very thin plates for CSR cladding, is the main aim of the present study. Making the absorber rod thinner and CSR cladding thicker by introducing a better blackness absorbing material and a new stainless steel alloy, respectively, are two studied ways to reduce the effects of mentioned problems.     

A study of the decommissioning procedure of an activated structure through an evaluation of the decommissioning cost for a research reactor

As decommissioning of a research reactor and a nuclear installation requires a long period of time from the decommissioning preparation work to the site remediation, the management of the data generated during the entire period of decommissioning is one of the most important tasks. In particular, the data obtained from research reactor decontamination and decommissioning activities can be important resources securing the safety and economic feasibility for other research reactor decommissioning. The owner of the research reactor and nuclear power plant need to submit decommissioning plan to the regulatory body at the starting stage of the research reactor and nuclear installation decommissioning project. The cost plan for decommissioning and the method for assessing the amount of exposure to protect workers must be stated in the decommissioning plan.This paper introduces the DES (Decommissioning Engineering System) that can be able to manage the data generated in the process of decommissioning of the TRIGA research reactor, to calculate an amount of waste, to evaluate decommissioning cost after deriving unit work productivity factor, and to predict the decommissioning process in advance. To verify the usability of this system and data integrity through connections among the unit systems, it describes the process to calculate the decommissioning cost using the data generated in dismantling an activated bio-shielding concrete in the TRIGA research reactor.As a result of the experiment to calculate the decommissioning cost with the TRIGA research reactor structure, it was found that the calculations were done precisely without flaw as the purpose of the experiment. Therefore, the DES can not only be used for other research reactors decommissioning, but also it is expected to be applied to other research reactors in the future. As a decommissioning cost between an activated concrete and a non-activated concrete according to the method of the dismantling procedure was significantly different, a study regarding the dismantling procedure needs more research.     

西安脉冲堆方波运行功率调节方法研究

本文建立了数字化功率调节系统在方波运行下的自动控制方法,提出了新的功率调节系统投入时刻设计方案,给出了调节棒初始棒位和发射脉冲棒后堆芯正反应性大小的建议,设计了数字化系统的方波运行功率调节方法,使用西安脉冲堆仿真程序XPRSC对设计方法进行了优化和验证。结果表明,该方法可实现更宽定值功率范围内的方波运行,能为堆上试验提供理论指导。     

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.     

气动式脉冲液体射流泵性能实验研究

实验研究了喷嘴直径为5 mm,扩散管直径分别为5、7 mm,提升高度为6.7 m的气动式脉冲液体射流泵的性能。结果表明:料桶内的液面高度对气动式脉冲液体射流泵的效率、扬程、输送量影响甚微,而随着操作压力的增加,脉冲液体射流泵的效率、扬程、输送量也增大。证明了在本实验条件下,最佳喷嘴间距与喷嘴直径之比为0.8～2.0范围内,并讨论了不同喷嘴间距所对应的最低操作压力条件。     

强流脉冲离子束产生的实验研究

分析了箍缩反射离子的工作原理，实验研究了二极管结构参数对其电参数的影响。在200kV加速器装置上获得了最大峰值2.lkA、脉宽约20ns、束能5．5J的离子流，最大离子产生效率达到6.2％。     

基于GDT的聚变裂变混合堆堆芯参数初步设计研究

基于Gas Dynamic Trap(GDT)装置的实验进展,提出了用于驱动聚变裂变混合堆包层的聚变堆芯参数设计。基于零维堆芯物理模型,计算分析给出了一套聚变功率为50MW的初步堆芯参数方案。利用GDT装置的实验结果对该物理模型进行计算对比校验,显示该物理模型和设计参数的可靠性。     

Design study on reactor structure of Pb–Bi-cooled direct contact boiling water fast reactor (PBWFR)

Pb–Bi-cooled direct contact boiling water small fast reactor (PBWFR) can produce steam by direct contact of feedwater with primary Pb–Bi coolant above the core, and circulate Pb–Bi coolant by means of buoyancy of steam bubbles. The PBWFR is capable of eliminating components of the cooling system such as primary pumps and steam generators, and thereby making the reactor system simple and compact. The specifications of the PBWFR are as follows: the fuel is Pu–U nitride; the core height is 75 cm; the core diameter is 278 cm; the average burnup is 80 GWd/t; the refueling interval is 10 years; the rated electric power is 150 MWe; the rated thermal power is 450 MWt; the core outlet/inlet temperatures are 460 °C/310 °C, respectively; and the operating steam pressure is 7 MPa. The reactor structure design has been formulated, where reactor vessel sizes are 4200 mm (ID) × 19,750 mm (H), the guard vessel is a closed type, the upper structure is made of chimneys, and the core support structure is hung up. An ultrasonic flow meter is installed inside the vessel. The seismic evaluation, design of refueling procedure and cost evaluation have been performed.