低中水平放射性废物浅地层处置安全评价方法研究

中国辐射防护研究院和日本原子力研究所合作开展了为期５年（１９９８年１月１９９３年１月）的“低中水平放射性废物浅地层处置安全评价方法研究”,以建立一套低中放废物浅地层处置安全评价技术和方法,包括参数、模式和程序。本文主要介绍在黄土包气带中核素迁移规律,水分运行行为研究及其相关参数测定的方法和试验结果。以及试验场址主要特征和开发的素迁移式和计算程序。为开展现场核素迁移 实验室核素迁移模拟实验,建立了野     

溶度积约束—废物处置安全评价中不应忽视的问题

对废物处置安全评价中常用的核素迁移计算方法进行了分析。利用溶度积约束对废物源项进行修正,提出了一个计算核素在地下水中迁移的方法。     

PWR核电厂乏燃料贮存临界计算重要核素的选取

信用核素选取是基于燃耗信用制乏燃料贮存临界安全分析的关键一步。通过对不同富集度、燃耗深度及停堆冷却时间下典型PWR燃料组件分析,以核素中子吸收份额大小排序为依据,筛选出对总的中子吸收起主要贡献的核素。结果显示,47个核素即可包络停堆后0~20a内影响乏燃料贮存系统反应性的所有核素中的99%。通过核素敏感性因子分析证明依据中子吸收份额排序选取重要核素的方法是合理的,与基准算例的结果对比证明所筛选出的核素能足够代表影响系统反应性的所有重要核素。     

芬兰乏燃料最终地质处置研究历程及经验

对芬兰乏燃料地质处置库的选址、场址评价、地下实验室研究历程进行了介绍。芬兰在通过参与国际合作的基础上,提出了具本国特色的乏燃料地质处置技术路线,即“选址-特定场址地下实验室-处置库”方案。借鉴芬兰在乏燃料地质处置研究中积累的经验对我国高放废物处置场址的选择及其研究技术,作者提出了一些建议和具体的做法。     

基于SCALE的压水堆乏燃料贮存水池临界安全分析

基于国际先进的核设计与安全分析计算程序SCALE,针对我国自主研发的先进压水堆乏燃料贮存水池,建立恰当的计算模型,并选取合理的保守假设,计算乏燃料水池正常贮存及事故工况下的反应性,评估计算模型的临界安全,分析该程序对我国先进反应堆乏池计算的适用性。计算结果表明该先进压水堆乏燃料贮存水池正常贮存工况及事故工况的有效增值因子均小于0.95,处于次临界状态。该设计模型可确保燃料堆内贮存区域临界状态安全可控。SCALE计算程序适用于我国自主研发的先进压水堆乏燃料水池临界安全计算。     

非能动压水堆核电厂乏燃料池风险评价

以非能动压水堆核电厂为研究对象,对可能引起乏燃料损伤的内部事件进行了风险评价。采用PSA软件RiskSpectrum建立事件树和故障树模型,进行乏燃料损伤频率（FDF）定量化。结果表明：在所有工况下总的FDF为2.05×10^-9/(堆•年),远小于堆芯的损伤频率（约2.41×10^-7/(堆•年)）;即使在放射性完全释放的假设下,乏燃料损伤导致的大量放射性释放频率仍较堆芯损伤导致的大量放射性释放频率（约2.38×10^-8/(堆•年)）低1个量级;由于非能动压水堆核电厂有多重预防缓解措施以应对乏燃料池（SFP）事故,SFP风险远低于堆芯风险,可实现核安全导则的安全目标。     

黄土包气带中放射核素迁移的现场试验

在中国辐射防护研究院野外试验场琥展了为期二年的包气带中＾３Ｈ、＾６０Ｃｏ、＾８５Ｓｒ和＾１３４Ｃｓ的迁移试验,在天然条件和人工喷啉条件下共进行了六组试验。通过期取样和直接测量两种方法测量了示踪核素的浓度分布;同时为配合迁移试验,同步开展了包气带中水分运行观测。     

乏燃料溶液系统临界安全分析计算方法研究

核临界安全分析是保证乏燃料后处理厂安全性的关键技术,而现有核临界安全事故分析程序中,或在几何适用范围上受限,或由于计算效率低而工程实用性差。因此,亟需研发一套适用范围大、计算精度高的临界安全分析方法,提高对核临界事故的分析精度,为乏燃料后处理厂提供技术保障。为此,本文针对乏燃料溶液系统特性,基于零维超细群截面制作与全问题并群方法、预估-校正准静态中子动力学计算方法和二维轴对称热工-辐解气体模型,开发了相应的计算程序模块,最终形成了一套具备并行功能的三维乏燃料溶液系统临界安全分析程序hydra-TD。进一步利用该程序对法国SILENE实验装置进行了验证,结果显示：第一裂变功率峰、倍增时间、总裂变次数等关键参数的误差较小,证明hydra-TD程序正确模拟了燃料溶液系统临界过程中的多物理过程,具备临界安全分析的能力。     

核电站乏燃料贮存水池失去最终热阱时的安全分析

压水堆核电站一回路和乏燃料贮存水池的设备冷却水由海水冷却器提供.本文假设事故工况下,海水冷却器突然停止工作,利用热平衡方程,计算并分析了乏燃料贮存水池运行的安全性及作为冷却水源冷却其它一回路重要用户的可能性.计算表明:在本文的各种工况下,乏燃料贮存水池运行是安全的;除一种工况外,硼水还具有冷却其它设备的能力.     

树立对深地质处置安全的信心

【英国《国际核工程》2002年8月刊报道】 深地质处置库的建造工作需要分阶段进行,并耗费数十年时间。核能协会(NEA)发表了一份题为“树立和交流对深地质处置安全的信心”的报告。 NEA性能评估咨询组于1994年组建了深地质处置库综合性能评估工作组(IPAG),主要目的是讨论安全性、评估性能、全面检查安全案例及其支持的综合性能评估(IPA)研究。第三个工作组,即IPAG-3是评价用于建立安全信心的方法。IPAG-3的目标是评价获得、表述和展示长期安全信心的技术状况,提出未来如何引导和创新以增强信心的建议。 建立深地质处置库需要分阶段,…     

Analyses of disposal efficiency based on nuclear spent fuel cooling time and disposal tunnel/pit spacing for the design of a geological repository

The purpose of deep geological disposal of high-level radioactive waste (HLW) including nuclear spent fuels is to isolate and to inhibit the release of radioactive material for a long time so that its toxicity does not affect the biosphere. The main requirement for the HLW repository design is to keep the buffer temperature below 100 °C in order to maintain the integrity of the engineered barrier system. The cooling time of the spent fuels discharged from nuclear power plants is the key consideration factor for the efficiency and economic feasibility of such a repository. We analyze the spacing of the disposal tunnels and pits, the disposal area and the uranium density for the deep geological repository layout to satisfy the thermal requirement of the disposal system. To do this, thermal stability analyses of a disposal system have been performed using varying spent fuel cooling times and spacing of the disposal tunnels and pits. The results show that the time to reach the maximum temperature within the design limit of the temperature in the disposal site is likely to be shortened as the cooling time of the spent fuel becomes shorter. Also it seems that controlling the disposal pit spacing is considered more advantageous than controlling the disposal tunnel spacing to meet the allowable thermal criteria in the repository from thermal and economical points of view. The results of these analyses can be used for a deep geological repository design and detailed analyses with exact site characteristics data will reduce the uncertainty of the results.     

聚变裂变混合乏燃料焚烧堆FDS-SFB燃料循环方案设计与分析

针对聚变裂变混合乏燃料焚烧堆FDS-SFB(Spent Fuel Burner),基于湿法和干法两种后处理技术途径提出了不同的燃料循环方案。并分别对FDS-SFB燃料循环所需的初装资源量、燃料制备和乏燃料后处理能力进行初步质量流分析和可行性初步评估。基于较好嬗变和增殖性能的FDS-SFB典型中子学方案的质量流初步分析表明:两种方案燃料循环其所需的初装资源量、燃料制备、乏燃料后处理能力具有初步的可行性。     

加拿大乏燃料处置的研发工作

论述了加拿大从1974年至今的乏燃料地质处置的研发工作,并对2002年以后研发工作中的某些问题进行了较详细的介绍:加拿大核废物管理机构、加拿大第4个乏燃料处置方案——可调整的分期管理(APM)方案的产生、APM方案的特点及其实施的技术路线,以及今后5 a(2011～2015年)的战略计划.     

A new assessment method for demonstrating the sufficiency of the safety assessment and the safety margins of the geological disposal system

A new method for demonstrating the sufficiency of the safety assessment and safety margins of the geological disposal system has been developed. The method is based on an existing comprehensive sensitivity analysis method and can systematically identify the successful conditions, under which the dose rate does not exceed specified safety criteria, using analytical solutions for nuclide migration and the results of a statistical analysis. The successful conditions were identified using three major variables. Furthermore, the successful conditions at the level of factors or parameters were obtained using relational equations between the variables and the factors or parameters making up these variables. In this study, the method was applied to the safety assessment of the geological disposal of transuranic waste in Japan. Based on the system response characteristics obtained from analytical solutions and on the successful conditions, the classification of the analytical conditions, the sufficiency of the safety assessment and the safety margins of the disposal system were then demonstrated. A new assessment procedure incorporating this method into the existing safety assessment approach is proposed in this study. Using this procedure, it is possible to conduct a series of safety assessment activities in a logical manner.     

Modeling spent TRISO fuel for geological disposal: corrosion and failure under oxidizing conditions in the presence of water

High temperature gas reactors (HTGRs) are being considered for near term deployment in the United States under the GNEP program and farther term deployment under the Gen IV reactor design (U.S. DOE Nuclear Energy Research Advisory Committee, 2002). A common factor among current HTGR (prismatic or pebble) designs is the use of TRISO coated particle fuel. TRISO refers to the three types of coating layers (pyrolytic carbon, porous carbon, and silicon carbide) around the fuel kernel, which is both protected and contained by the layers. While there have been a number of reactors operated with coated particle fuel, and extensive amount of research has gone into designing new HTGRs, little work has been done on modeling and analysing the degradation rates of spent TRISO fuel for permanent geological disposal. An integral part of developing a spent fuel degradation modeling was to analyze the waste form without taking any consideration for engineering barriers. A basic model was developed to simulate the time to failure of spent TRISO fuel in a repository environment. Preliminary verification of the model was performed with comparison to output from a proprietary model called GARGOYLE that was also used to model degradation rates of TRISO fuel. A sensitivity study was performed to determine which fuel and repository parameters had the most significant effect on the predicted time to fuel particle failure. Results of the analysis indicate corrosion rates and thicknesses of the outer pyrolytic carbon and silicon carbide layers, along with the time dependent temperature of the spent fuel in the repository environment, have a significant effect on the time to particle failure. The thicknesses of the kernel, buffer, and IPyC layers along with the strength of the SiC layer and the pressure in the TRISO particle did not significantly alter the results from the model. It can be concluded that a better understanding of the corrosion rates of the OPyC and SiC layers, along with increasing the quality control of the OPyC and SiC layer thicknesses, can significantly reduce uncertainty in estimates of the time to failure of spent TRISO fuel in a repository environment.     

放射性废物处置安全全过程系统分析中的指标

以往放射性废物处置的安全评价中通常使用放射性安全指标(即剂量和危险),随着放射性废物处置安全全过程系统分析这一概念的提出,辅助指标已成为评价中的一个重要组成部分。本文介绍了安全全过程系统分析中所使用指标的发展、分类和相应标准等。依评价对象不同,辅助指标通常分为安全指标和性能指标,有些组织还提出了安全功能指标;与上述指标相对应的用于比较的标准分别为参考值、指标标准和安全功能指标标准。将处置系统划分为不同库室时,指标还可分为“包容物和浓度”相关指标、“通量”相关指标和“屏障状态”相关指标三类。建议我国尽快开展放射性废物处置的安全全过程系统分析工作,建立完善的指标体系,选取适当的评价指标,并基于我国放射性废物处置的场址特性确定相应的标准,以期实现安全和防护的最优化。     

The effect of organic retarders on grout thickening and setting during deep borehole disposal of high-level radioactive waste

Deep borehole disposal (DBD) is being increasingly seen as a viable and potentially superior alternative to comparatively shallow mined repository concepts for disposal of some high-level radioactive wastes. We report here details of proof-of-concept investigations into the use of cementitious grouts as sealing/support matrices for use in low temperature DBD scenarios. Using the cementitious grout to fill annular space within the disposal zone will not only support waste packages during placement, but will also provide a low permeability layer around them which will ultimately enhance the safety case for DBD. Grouts based on Class G oil well cement are being developed. The use of retarders to delay the accelerated onset of thickening and setting (caused by the high temperature and pressure in the borehole) is being investigated experimentally. Sodium gluconate and a polycarboxylate additive each provide sufficient retardation over the range 90–140 °C in order to be considered for this application. Phosphonate and sulphonate additives provide desirable retardation at 90 °C. The additives did not affect grout composition at 14 days curing and the phases formed are durable at elevated temperature and pressure.     

The adsorption and desorption of water from a carbonaceous deposit layer on the surface of stainless steel representing spent AGR nuclear fuel cladding

Spent fuel discharged from advanced gas-cooled reactor power stations carries a deposit of carbon firmly attached to the cladding surface. The fuel route involves contact with water, for cooling and transport. Long-term storage potentially includes dry storage, however, the carry-over of water entrained within the carbon deposit needs to be considered regarding the storage environment. Drying of the fuel is possible, but little is known concerning the drying characteristics of such deposits. This work reports preparation of a laboratory simulant of a carbon deposit on a fuel pin surface and measurement of its adsorption and desorption properties regarding liquid and vapour phase water. This work found that water vapour equilibration is rapid and reversible. Liquid water uptake is appreciable (up to 5.7 times the mass of carbon) and most (up to 88%) is removed on standing for 12 h. Heating removes little more. The implications for spent fuel management are discussed.     

低中水平放射性固体废物近地表处置设施安全分析报告格式和内容的探讨

我国目前尚缺少放射性废物处置安全分析报告格式、内容要求的相关导则,本文在基于我国已有低中放固体废物近地表处置工程实践及国际原子能机构安全全过程系统分析理念的基础上,对近地表处置安全分析报告的编制要求进行探讨,提出对应的格式与内容建议,以期为编制低中水平放射性固体废物近地表处置设施安全分析报告格式和内容给的标准提供参考。