Corrosion assessment for spent nuclear fuel disposal in crystalline rock,using variant cases of hydrogeological modeling

This paper presents a corrosion assessment of copper spent nuclear fuel disposal canisters in crystalline rock,using hydrogeological modeling.A simplified approach is considered,to avoid complex and time-consuming computer simulations.This simplified case is presented as a base case,with changes in the hydrogeological parameters presented as variant cases.The results show that in Taiwan’s base case,decreasing the hydraulic conductivity of the rock or decreasing the hydraulic conductivity of dikes results in a shorter transport path for sulfide and an increase in corrosion depth.However,the estimated canister failure time is still over one million years in the variant cases.

Seismic considerations for spent nuclear fuel storage in dry casks

To aid the United States Nuclear Regulatory Commission,Sandia National Laboratories (SNL) was contracted to investigate the seismic behavior of typical dry cask storage systems. Parametric evaluations characterized the sensitivity of calculated cask response characteristics to input parameters. The parametric evaluation investigated two generic cask designs (cylindrical and rectangular),three different foundation types (soft soil,hard soil,and rock),and three different casks to pad coefficients of friction (0.2,0.55,0.8) for earthquakes with peak ground accelerations of 0.25g,0.6g,1.0g and 1.25g. A total of 1 165 analyses were completed,with regression analyses being performed on the resulting cask response data to determine relationships relating the mean (16 % and 84 % confidence intervals on the mean) to peak ground acceleration,peak ground velocity,and pseudo-spectral acceleration at 1 Hz and 5 % damping. In general,the cylindrical casks experienced significantly larger responses in comparison to the rectangular cask. The cylindrical cask experienced larger top of cask displacements,larger cask rotations (rocking),and more occurrences of cask toppling (the rectangular cask never toppled over). The cylindrical cask was also susceptible to rolling once rocking had been initiated,a behavior not observed in the rectangular cask. Cask response was not overly sensitive to foundation type,but was significantly dependent on the response spectrum employed.

Challenges in spent nuclear fuel final disposal:conceptual design models

<正>The disposal of spent nuclear fuel is a long-standing issue in nuclear technology.Mainly,UO_2 and metallic U arc used as a fuel in nuclear reactors.Spent nuclear fuel contains fission products and transuranium elements,which would remain radioactive for 10~4 to 10~8 years.In this brief communication,essential concepts and engineering elements related to high-level nuclear waste disposal are described.Conceptual design models are described and discussed considering the long-time scale activity of spent nuclear fuel or high level waste.Notions of physical and chemical barriers to contain nuclear waste are highiightened.Concerns regarding integrity,self-irradiation induced decomposition and thermal effects of decay heat on the spent nuclear fuel are also discussed.The question of retrievability of spent nuclear fuel after disposal is considered.

Safe Controlled Storage of SVBR-100 Spent Nuclear Fuel in the Extended-Range Future

Experience of operating reactor facilities (RF) with lead-bismuth coolant (LBC) has revealed that it is possible to perform safe refueling in short terms if the whole core is replaced and a kit of the special refueling equipment is used. However, comparing with RFs of nuclear submarines (NS), in which at the moment of performance of refueling the residual heat release is small, at RF SVBR-100 in a month after the reactor has been shut down, at the moment of performance of refueling the residual heat release is about 500 kW. Therefore, it is required to place the spent removable unit (SRU) with spent fuel subassemblies (SFSA) into the temporal storage tank (TST) filled with liquid LBC, in which the conditions for coolant natural circulation (NC) and heat removal via the tank vessel to the water cooling system are provided. After the residual heat release has been lowered to the level allowing transportation of the TST with SRU in the transporting-package container (TPC), it is proposed to consider a variant of TPCs transportation to the special site. On that site after the SRU has been reloaded into the long storage tank (LST) filled with quickly solidifying liquid lead, the TPCs can be stored during the necessary period. Thus, the controlled storage of LSTs is realized during several decades untill the time when SNF reprocessing and NFC closing are becoming economically expedient. On that storage, the four safety barriers are formed on the way of the release of radioactive products into the environment, namely: fuel matrix, fuel element cladding, solid lead and steel casing of the LST.

Numerical simulation of coupling heat transfer and thermal stress for spent fuel dry storage cask with different power distribution and tilt angles

Dry storage containers must be secured and reliable during long-term storage,and the effect of decay heat released from the internal spent fuel on the cask has become an important research topic.In this paper,a 3D computational fluid dynamics model is presented,and the accuracy of the calculation is verified,with computational errors of less than 6.2%.The thermal stress of the dry storage cask was estimated by coupling it with a transient temperature field.The total power remained constant and adjusting the power ratio of the inner and outer zones had a small effect on the stress results,with a maximum equivalent stress of approximately 5.2 kPa,which occurred at the lower edge of the shell.In the case of tilt,the temperature gradient varied in a wavy distribution,and the wave crest moved from right to left.Altering the tilt angle affects the air distribution in the annular gap,leading to the shell temperature being transformed,with a maximum equivalent stress of 202 MPa at the bottom of the shell.However,the equivalent stress in both cases was less than the yield stress(205 MPa).

Systematic impact of spent nuclear fuel on θ_(13) sensitivity at reactor neutrino experiment

Reactor neutrino oscillation experiments, such as Daya Bay, Double Chooz and RENO are designed to determine the neutrino mixing angle θ13 with a sensitivity of 0.01--0.03 in sin^2 2θ13 at 90% confidence level, an improvement over the current limit by more than one order of magnitude. The control of systematic uncertainties is critical to achieving the sin^22θ13 sensitivity goal of these experiments. Antineutrinos emitted from spent nuclear fuel （SNF） would distort the soft part of energy spectrum and may introduce a non-negligible systematic uncertainty. In this article, a detailed calculation of SNF neutrinos is performed taking account of the operation＂ of a typical reactor and the event rate in the detector is obtained. A further estimation shows that the event rate contribution of SNF neutrinos is less than 0.2% relative to the reactor neutrino signals. A global X2 analysis shows that this uncertainty will degrade the θ13 sensitivity at a negligible level.

A study of antineutrino spectra from spent nuclear fuel at Daya Bay

The Daya Bay Reactor Antineutrino Experiment is designed to determine the as yet unknown neutrino mixing angle,θ13,by measuring the disappearance of electron antineutrinos from several nuclear reactor cores.The projected sensitivity in sin2（2θ13） of better than 0.01 at a 90% CL should be achieved after three years of data-taking.Antineutrinos emitted from spent nuclear fuel （SNF） distort the soft part of the energy spectrum.In this article,a calculation of the antineutrino spectra from the long-life isotopes in SNF is performed.A non-equilibrium generation of long half-life isotopes during the running time of the reactor is also analyzed.Finally,we show that the antineutrino event rate contribution from SNF,which has been stored in the SNF pool for several years,may be non-negligible.

Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR/CANDU synergism

High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (～22.5%), increase the energy output (～41%), decrease the quantity of spent fuels to be disposed (～2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.

Thermodynamic Assessment of UO₂Pellet Oxidation in Mixture Atmospheres under Spent Fuel Pool Accident

For an analysis of the oxidation behavior of UO2 nuclear fuel pellet under a loss of water coolant accident in a spent nuclear fuel pool of an LWR, thermodynamic assessments of UO2 oxidation were carried out under various atmospheric conditions. In a steam atmosphere, it was assessed that UO2 would not be fully oxidized into U3O8 due to the relatively lower oxygen partial pressure, while UO2 will be fully oxidized into U3O8 in an air atmosphere. In an air and steam mixture atmosphere, the UO2 oxidation was dominantly affected by the air volumetric fraction, because of the relatively higher oxygen partial pressure of air. In addition, the effect of H2 volumetric fraction on the oxygen partial pressure under a mixture atmosphere was calculated, and it was revealed that UO2 pellet oxidation could be reduced above the critical value of H2 volumetric fraction.

乏燃料后处理碱性流程的研究进展

乏燃料后处理碱性流程是用碳酸盐、氢氧化物等碱性物质的溶液作为介质进行乏燃料的溶解及铀、钚等元素的分离与纯化的方法。碱性条件下,乏燃料中的大部分裂变产物和次锕系元素并不溶解或者在溶解过程中转变为碳酸盐、氢氧化物沉淀。与已经实现工业化的PUREX(plutonium uranium redox extraction)酸性流程相比,碱性流程具有腐蚀性更小、流程更简单等潜在的优点。鉴于碱性流程的优点及其在乏燃料后处理中的潜在应用,日本、美国、俄罗斯、韩国等国家的科研人员已经围绕该流程开展了一些研究工作。本文首先介绍了各国建议的碱性流程的技术路线;然后逐一介绍了与主要工艺环节相关的基础研究的进展,包括乏燃料的氧化溶解、核素分离、试剂的回收等;最后对该领域面临的挑战和前景进行了讨论。

基于D-D中子源的乏燃料组件钚含量测量装置设计

为了满足国际社会上对于乏燃料组件内特种可裂变材料钚的保障监督需求,以持续提供核材料信息,研究乏燃料组件内钚含量测量的非破坏性分析技术显得尤为重要。本文采用主动法,以压水堆乏燃料组件作为测量对象,D-D中子作为质询中子源,开展了乏燃料组件钚含量测量装置的设计研究。采用MCNPX软件,基于最大化探测器计数率和使各探测器计数率尽量一致的目的,对测量装置中的中子管与探测器组件距离、中子管慢化材料及其厚度、探测器组件与中子管高度差、探测器组件中慢化体厚度等关键参数进行了模拟计算。此研究为中子质询乏燃料组件钚含量测量技术研究及实验验证打下了基础。

国外乏燃料干法后处理设施进展

干法后处理技术具有介质耐辐照、临界风险低、工艺流程短、废物量小等特点,是核燃料后处理领域中适应性更高、处理对象更广的一种分离技术。干法后处理设施是实现干法后处理技术开发、验证和应用的关键场所。本文调研总结了国外干法后处理技术研发和示范设施进展,从设施建设背景、工艺基准流程、主要技术参数、设施布局设计和应用情况等多方面进行了分析和比较,并结合我国干法后处理技术发展现状和设想,提出了我国干法后处理设施发展建议。

压水堆乏燃料组件中主要同位素含量分析计算

绝大多数的特种可裂变材料钚均存在于商用乏燃料组件中,在核保障领域,对于乏燃料组件中钚含量的核实显得尤为重要。为分析计算乏燃料组件中易裂变核素钚的含量,以31组AFA-2G典型压水堆乏燃料组件为研究对象,通过分析核燃料在反应堆运行过程中的U-Pu循环燃耗链的特点,分析计算了乏燃料组件中主要同位素的含量和固有中子本底,同时利用蒙特卡罗模拟方法计算了239Pu等效质量的等效系数。本文工作为中子质询乏燃料组件易裂变核素含量非破坏性测量的方法研究及测量装置的设计打下了基础。

混合式K边界/X荧光密度计测铀钚质量浓度的不确定度分析

为检验混合式K边界技术测量混合溶液中铀钚质量浓度的精度,使用混合式K边界/X荧光密度计(hybrid K-edge/XRF densitometer,HKED)对不同混合溶液进行了铀、钚质量浓度测量实验。根据实验数据计算了测量结果的不确定度,并分析了测量过程中的偏差来源。通过对测量结果及不确定度的分析,验证了混合式K边界技术测量铀、钚质量浓度可达到核安全导则《核燃料后处理厂核材料衡算》中的测量精度,对标定范围内铀质量浓度的测量不确定度小于0.28%,钚质量浓度的测量不确定度基本小于0.94%,精度较高,能够满足生产中的衡算要求;总结了测量中的主要偏差来源和优化方向,为进一步提高测量精度和拓展混合式K边界技术在锕系元素测量中的应用提供数据支持。

水溶性三价镧锕分离配体研究进展

三价镧系元素和次锕系元素的分离是核工业中乏燃料处理的关键。为实现两者的有效分离,通过合理的配体结构设计从而选择性的络合一种金属进而达到分离的目的是当前液-液萃取配体设计的通用策略。其中,高效水溶性配体的设计能够极大程度的降低液-液萃取过程中有机溶剂的使用进而备受关注。本文简要的梳理了近10年来水溶性配体的发展,根据致溶基团将水溶性镧锕分离配体分为三个部分并讨论了不同结构配体的优缺点。最后结合我们课题组的近期工作进展对水溶性配体的发展方向进行了展望。

高燃耗乏燃料干贮系统设计与验证研究

随着国内核电的快速发展,大量的乏燃料从反应堆中卸出,乏燃料的贮存问题具有迫切性与必要性。对高燃耗乏燃料贮存方法进行理论研究,设计了基于强制氮循环系统的乏燃料干式贮存综合台架。该系统包括冷凝、去湿、氮循环、密封容器四个模块,通过氮气循环对高燃耗乏燃料进行干燥与降温。本文对该系统进行模拟仿真验证与台架试验验证,利用仿真结果与试验结果进行相互印证,试验结果表明:该台架符合设计预期,能够在设计时间内通过氮循环系统对密封容器内部起到干燥及冷却的效果,为高燃耗乏燃料贮存提出了可行方法。

多元统计方法在核取证溯源分析中的研究进展

国家高度重视核安全,习近平总书记多次在全球核安全峰会上强调核材料溯源取证的重要性。核材料溯源取证,是对核材料及放射性材料进行检查分析,以溯源其性质、制造的时间、地点、方式及预期用途。目前关于该领域的前沿研究方向是基于数据库信息,利用多元统计方法溯源未知核材料的来源信息。该文回顾了近年来的核取证研究进展,并指出存在数据依赖性强、模型缺乏定量分析能力和普适性等不足。为克服这些局限,该文介绍了最新的线性拟合方法实现溯源取证分析。基于经济合作与发展组织核能署(OECD/NEA)于2017年发布的SFCOMP-2.0乏燃料数据库,以不同核素浓度作为样品特征,提出了线性关系假设并通过数据库数据和压水堆(PWR)与沸水堆(BWR)模拟结果进行了检验。结果表明,线性关系假设在PWR和BWR中均适用,并可用于预测燃料的初始富集度和燃耗量。此外,应用3种机器学习方法(逻辑回归、支持向量机、多层感知器)实现对PWR和BWR的分类。最后,应用KNN分类、随机森林和多层感知器算法改进了核取证分类模型,提高了模型对各反应堆的分辨能力。

秦山核电站海域有害盐在带温核级材料表面沉积实验设计

含Cl有害盐在服役构件表面的沉积量,是影响服役构件腐蚀进程的重要因素。秦山核电站临海而建,面临含Cl有害盐沉积引起的腐蚀问题。该文通过实地环境调研,并根据调研结果开展实验室盐雾沉积实验设计,结果表明:以ASTM D1141-98(2021版)标准中的人工海水Cl元素含量为基准,将秦山核电站海域海水各化合物含量乘以4.075 26进行放大,可得到符合ASTM标准设计的盐雾沉积用有害盐模拟溶液成分。之后,该文进一步开展90℃带温核级材料表面有害盐沉积实验,探究临海服役环境下,有害盐在秦山核电站乏燃料贮罐材料(带温核级材料)表面的沉积规律,为开展实际服役环境下的核电站材料服役寿命评估、服役性能评价提供一种实验室设计思路与借鉴。

中美核燃料循环设施核事故应急状态分级对比与探讨

本文介绍了美国核管会(NRC)及中国核燃料循环设施应急状态分级发展的历史及现状,对比了中美核燃料循环设施应急状态分级的差异,并给出分析结果,建议对后处理设施开展完整的二级PSA研究,给出相关事故谱,为进行应急状态分级及应急行动水平制定提供充分的技术支撑。

乏燃料后处理厂核应急评价与决策支持系统设计

乏燃料后处理厂可能发生临界、放射性物质泄漏、火灾和爆炸等事故,营运单位需要建立相应的应急评价能力,配置针对上述事故的核应急评价系统。本文介绍了针对乏燃料后处理厂5种典型事故的三维可视化实时核应急评价与决策支持系统设计,该系统可基于工艺系统监测数据实现应急工况实时诊断,计算向厂房和环境释放的源项,基于应急预案开展应急响应流程管理,针对工作人员和公众防护策略开展防护行动分析等功能,并基于三维可视化技术实现应急评价结果和响应流程的动态展示。本系统可用于我国乏燃料后处理厂应急评价与决策支持,提升其应急准备与响应能力。