Safety analysis of beyond design basis accidents in RBMK-1500 reactors

At present the design basis accidents for RBMK-1500 are rather thoroughly investigated. The performed analyses helped to develop and implement a number of safety modifications. Further plant safety enhancement requires developing emergency procedures that would enable beyond design basis accidents management by preventing core damage or mitigating consequences of severe accidents.     

核设施退役辐射检测及其仪器仪表应用

核设施退役辐射场检测技术在核设施退役的整个过程中起着非常重要的作用.较系统论述了核设施退役辐射检测的质量保证,最后从满足核设施退役工作的需要出发,评述了退役过程辐射检测仪器仪表的选择及其应用.指出国内辐射监测仪器仪表在满足核设施退役的需求上的较大差距,值得各有关方面关注.     

国内压水堆核电厂安全壳喷淋试验的比较分析

以秦山第二核电厂1、2号机组和田湾核电厂1、2号机组为例,对国内已运行的压水堆核电厂在调试阶段进行安全壳喷淋试验时所采用的两种方式进行了比较分析,总结出两种喷淋试验方法各自的优缺点,可为国内新建核电机组的安全壳喷淋系统的调试试验提供借鉴。     

Experimental investigation of the serviceability of protective cladding for microfuel under the conditions of serious accidents in light-water reactors

The serviceability (corrosion resistance and integrity) of microfuel with a protective coating consisting of silicon carbide is investigated for the conditions of serious loss-of-coolant accidents in light-water reactors. Tests in a synthetic steam-gas medium — products of the combustion of propane in oxygen — which were performed in the temperature range 730–1670°C established that at temperatures up to 1590°C the microfuel coating possesses high corrosion resistance and retains its integrity. Substantial damage (up to 25%) to the microfuel was observed at 1670°C. The microfuel successfully passed tests with heating up to 800–1600°C in air and subsequent rapid cooling in water. __________ Translated from Atomnaya énergiya, Vol. 103, No. 5, pp. 302–309, November, 2007.     

Enhanced tritium production for fusion reactors via3He(n,p)3 H in the heavy water moderator of a CANDU reactor

There is an increasing requirement for tritium to supply the fuel needs of current experimental fusion devices and in the initial startup of future power generating reactors. Tritium is produced in heavy water reactors through deuterium activation, but the total production capacity of Canadian operated CANDUs will fall short of future demands, during the period before and for some time after self-sufficient reactors become available. Consequently, methods of enhancing tritium generating rates warrant investigation. Herein we provide the results of an inquiry into the feasibility of enhancing tritium production levels through the activation of helium-3 following its external addition to the heavy water moderator system of a hypothetical 500–600 MWe CANDU reactor. The approach adopted involves simulation of the temporal evolution of the tritium activities, originating from²H(n,)³H and³He(n, p)³H, as described by a simple first order kinetic model. The results suggest that the frequent addition of helium-3 to the moderator water will enhance tritium production inventories. The enhancement factor is highly dependent upon the rate at which helium-3 irretrievably escapes to the moderator cover gas. However, the direct activation of helium-3, contained in a closed loop such as the annulus gas system, for example, would be essentially complete within a few weeks without any significant loss.     

Results of analyzing accidents with core meltdown in fast reactors with sodium as the coolant

Nizhegorod Polytechnical Institute. Translated from Atomnaya Énergiya, Vol. 73, No. 6, pp. 433-438, December, 1992.     

Neutronics shielding analysis for the end plug of a tandem mirror fusion reactor

A neutronics analysis using the Monte Carlo method is carried out for the end-plug penetration and magnet system of a tandem mirror fusion reactor. Detailed penetration and the magnets' three-dimensional configurations are modeled. A method of position dependent angular source biasing is developed to adequately sample the DT fusion source in the central cell region and obtain flux contributions at the penetration components.To assure cryogenic stability, the barrier cylindrical solenoid is identified as needing substantial shielding of about 1 m of a steel-lead-boron-carbide-water mixture. Heating rates there would require a thermal-hydraulic design similar to that in the central cell blanket region. The transition coils, however, need a minimal 0.2 m thickness shield. The leakage neutron flux at the direct converters is estimated at 1.3×10¹⁵ n/(m²·s), two orders of magnitude lower than that reported at the neutral beam injectors for tokamaks around 10¹⁷ n/(m²·s) for a 1 MW/m² 14 MeV neutron wall loading. This result is obtained through a coupling between the nuclear and plasma physics designs in which hydrogen ions rather than deuterium atoms are used for energy injection at the end plug, to avoid creating a neutron source there. This lower and controllable radiation leakage problem is perceived as a potential major advantage of tandem mirrors compared to tokamaks and laser reactor systems.     

Hydrogen management and overpressure protection of the containment for future boiling water reactors

New design and evaluation method for hydrogen management of containment atmosphere have been developed for application in the future boiling water reactor (BWR). These are intended as a part of consideration of severe accidents in the course of design so as to assure a high level of confidence that a large release of radioactivity to the environment that may result in unacceptable social consequences can reasonably be avoided. Emphasis on hydrogen management and protection against overpressure failure is based on the insights from probabilistic safety assessments (PSAs) that late phase overpressure (and associated leakage) and molten corium concrete reaction (MCCI) need attention to ensure that containment remains intact, in case energetic challenges to the containment such as DCH (direct containment heating) or FCI (fuel coolant interactions) are practically eliminated by design or resolved from risk standpoint of view. The authors studied the use of palladium-coated tantalum for hydrogen removal from containment atmosphere in order to avoid pressurization of the containment with small free volume by non-condensable gas and steam. Its effectiveness for ABWR (advanced boiling water reactor) containment was evaluated using laboratory test data. Although further experimental studies are necessary to confirm its effectiveness in real accident conditions, the design is a promising option and one that could be backfitted upon necessity to existing plants for which pressure retaining capability cannot be altered. Also new evaluation method for flammability control under severe accident conditions was developed. This method employes a realistic assessment of the amount of oxygen and hydrogen gases generated by radiolytic decomposition of water under severe accident conditions and their subsequent transport from water to containment atmosphere.     

谱估计方法在大型实验装置地基振动测量中的应用

大型物理实验装置为获得高品质的强流脉冲电子束流,通常对地基振动都有较高的要求。详尽的地质振动状况分析是装置选址的重要前提。本文针对某装置待勘区域实际地质条件和周围人文环境,首先简要介绍了测点布设、振动测试系统的基本组成,然后通过分阶段测试比对试验对某装置待勘区域进行实际振动测量,最后应用谱分析的方法对区域内不同时间段、不同方向的测量数据进行分析,评估选址位置是否合理。数据结果表明,待勘区域在5μm·s~(-1)以上速度被95%概率检测,垂直方向位移振幅谱在185 nm,水平方向大于垂直方向约60 nm,物理实验表明,进一步基建过程中应采取适当的减震措施降低垂直位移幅度。     

Evolution of the ITER program and prospect for the next-step fusion DEMO reactors: status of the fusion energy R&D as ultimate source of energy

An international joint project of fusion experimental reactor, the ITER (International Thermonuclear Experimental Reactor), is reviewed in view of long-range fusion energy research and development (R&D). Its purpose, goal, evolution, and the present construction status are briefly reviewed. While the ITER is a core machine in the present stage, generation of electricity is a role of the next-step fusion demonstration power plant “DEMO.” The status of designs and technology R&D for DEMO are also reviewed.     

Modeling of a confinement bypass accident with CONSEN,a fast-running code for safety analyses in fusion reactors

The CONSEN (CONServation of ENergy) code is a fast running code to simulate thermal-hydraulic transients, specifically developed for fusion reactors. In order to demonstrate CONSEN capabilities, the paper deals with the accident analysis of the magnet induced confinement bypass for ITER design 1996. During a plasma pulse, a poloidal field magnet experiences an over-voltage condition or an electrical insulation fault that results in two intense electrical arcs. It is assumed that this event produces two one square meters ruptures, resulting in a pathway that connects the interior of the vacuum vessel to the cryostat air space room. The rupture results also in a break of a single cooling channel within the wall of the vacuum vessel and a breach of the magnet cooling line, causing the blow down of a steam/water mixture in the vacuum vessel and in the cryostat and the release of 4 K helium into the cryostat. In the meantime, all the magnet coils are discharged through the magnet protection system actuation. This postulated event creates the simultaneous failure of two radioactive confinement barrier and it envelopes all type of smaller LOCAs into the cryostat. Ice formation on the cryogenic walls is also involved. The accident has been simulated with the CONSEN code up to 32 h. The accident evolution and the phenomena involved are discussed in the paper and the results are compared with available results obtained using the MELCOR code.     

乏燃料干式贮存设施临界计算研究

以HI-STORM 100乏燃料干式贮存设施内部装载AFA-3G燃料组件为研究对象,用MCNP(Monte Carlo N Particle Transport Code)4C程序,通过改变贮存设施内外的水密度,采用新燃料假设对不同工况下的临界安全进行研究。结果表明,在正常工况下,keff远低于0.93,是临界安全的。在事故工况下,当水密度大于0.8 g·cm-3时,存在临界安全问题。然后选取适当的核素,通过使用ORIGEN-ARP程序,得到不同燃耗下核素的组成,在同一模型下考虑燃耗信任制,对干式贮存设施的临界安全进行研究。在此基础上,给出了乏燃料干式贮存设施临界安全工作的相关建议。     

辐射监测气溶胶滤材过滤效率测试

采用双层滤膜采样法采集大气环境中的气溶胶,通过重量测量法、放射性核素测量法和特征元素测量法测量气溶胶采样滤膜对不同粒径颗粒物的过滤效率,利用这三种方法分别测试了两种常用滤膜的过滤效率。结果表明,不同类型滤膜的过滤效率存在明显差异;同一种滤膜对不同粒径颗粒物的过滤效率也有差别。     

VVER-1000堆型一回路水压试验准备与实施

描述了VVER-1000机组水压试验过程与步骤,结合上游设计文件、标准规范,通过与田湾一期、国内典型机组的比较,提出了试验验收准则、试验压力和温度要求、控制方法,总结了田湾3号机组初始一回路水压试验的准备与注意事项,对于水压试验程序的编制及试验的执行具有指导作用,可供后续同类型机组的水压试验和调试工作参考。     

The effect of wall loading limitations and choice of beta on the feasibility of advanced fuel fusion reactors

This paper investigates the effect of wall loading limitations and choice of plasma stability index beta on the feasibility of advanced fuel fusion reactors. Two new conceptual tools are introduced to facilitate this analysis: the “effective reactivity,” which includes all of the reaction-relevant parameters that determine the fusion power density, and the “critical radius,” which is the maximum allowable minor radius of a fusion reactor, beyond which the power generated in the plasma will exceed allowable loadings of radiant energy or neutrons on the first wall. It is shown that if high beta (greater than 0.2) fusion reactors are feasible, the high reactivity of the DT reaction cannot be fully exploited because of wall loading limitations. In addition, some high beta reactors with advanced fuels are also found to be wall loading limited, and to have excess reactivity, which can be traded off for lower magnetic fields, longer particle containment times, etc. Under certain circumstances, the reduced materials problems associated with some advanced fuels may outweigh the reactivity advantage of the DT reaction, and make one of them the reaction of choice for high beta fusion reactors.     

质量管理体系在核设备无损检测中的应用

通过对ISO 9001质量管理体系在核工业设备制造企业无损检测中的实际运行和应用,总结了在部门体系运行过程中的注意事项及特点,并针对质保审查中容易出现的问题提出了整改建议,从而提升了无损检测的体系运行状况。     

化学镀Ni-P技术在试验台架中的应用

在由碳钢制造的设备,管道和阀门等部位所组成的热工水力试验台架中,金属的氧化腐蚀是一个普遍存在的问题,本文简述了化学镀镍磷（Ni-P）工艺技术在试验台架建造中的并对试验体进行了验证,结果表明,该技术不仅能满足试验体对腐蚀杂质含量及颗粒度的严格要求,而且还可节省投资缩短项目建造周期,具有一定的工程应用价值。     

遗传算法在高通量工程试验堆燃料管理优化中的应用

建立了基于遗传算法的高通量工程试验堆(HFETR)堆内燃料管理优化模型。根据HFETR的实际情况,提出了基于组件位置的二进制编码/解码技术。研究了不同选择策略对算法性能的影响,探讨了遗传算法和专家经验的结合,吸收了自适应遗传算法的思想,得到了可直接应用于HFETR的优化换料方案。     

软X射线诊断冷却测试平台设计

软X射线物理诊断普遍应用于托卡马克聚变装置。在实验时,上限温度仅为75℃的软X射线诊断探测器将处于250℃的环境温度中,因此冷却系统不可或缺。本文分析了软X射线诊断探测器在运行时将受到的热负荷功率,并依据计算结果及实验时的冷却条件,创造性的设计了采用水、氦气混合制冷方式的冷却测试平台。所进行的测试结果表明,平台运行稳定,不仅能够很好的满足软X射线诊断的冷却需求,同时也为其它聚变诊断冷却系统的设计提供借鉴。