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.

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.

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.

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.

Physical modeling of spent-nuclearfuel container

A new physical simulation model was developed to simulate the casting process of the ductile iron heavy section spent-nuclear-fuel container.In this physical simulation model,a heating unit with DR24 Fe-Cr-Al heating wires was used to compensate the heat loss across the non-natural surfaces of the sample,and a precise and reliable casting temperature controlling/monitoring system was employed to ensure the thermal behavior of the simulated casting to be similar to the actual casting.Also,a mould system was designed,in which changeable mould materials can be used for both the outside and inside moulds for different applications.The casting test was carried out with the designed mould and the cooling curves of central and edge points at different isothermal planes of the casting were obtained.Results show that for most isothermal planes,the temperature control system can keep the temperature differences within 6 ℃ between the edge points and the corresponding center points,indicating that this new physical simulation model has high simulation accuracy,and the mould developed can be used for optimization of casting parameters of spent-nuclear-fuel container,such as composition of ductile iron,the pouring temperature,the selection of mould material and design of cooling system.In addition,to maintain the spheroidalization of the ductile iron,the force-chilling should be used for the current physical simulation to ensure the solidification of casting in less than 2 h.

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。

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。

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。

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。

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。

基于NECP-Bamboo程序的商用压水堆乏燃料组件核素成分分析

乏燃料组件核素成分的精确计算是乏燃料临界安全分析等工作的输入条件,放射性源项计算是乏燃料组件核素成分分析的典型应用。国内现有程序由于存在数据库中核素种类不全、辐照过程无法完全模拟等弊端,限制了乏燃料后处理安全分析的可靠性和经济性。本文基于完全自主化的压水堆堆芯分析软件NECP-Bamboo,研发了商用压水堆乏燃料组件核素成分计算程序Bamboo-SFuel,利用辐照后实验(PIE)实测数据对核素成分进行了定量验证与分析,通过与Scale程序包计算结果进行对比验证了程序源项计算的精度,还探究了不同燃耗数据库对核素成分和源项计算结果的影响。数值结果表明,Bamboo-SFuel能精确分析不同辐照条件下商用压水堆乏燃料组件的核素成分和放射性源项,使用NECP-Bamboo程序中不同核素数目的燃耗数据库对重要核素成分计算结果影响不大,但对总的放射性源项计算结果影响较大;基于内置的包含1547种核素的燃耗数据库,该程序可同时给出可靠的乏燃料临界安全分析和辐射安全分析关注的重要核素成分。

REMIX燃料可行性与乏燃料特性分析

为验证REMIX(回收铀钚混合氧化物)燃料在典型M310堆型上的适用情况并评价其源项和释热特性,使用CMS程序包研究了REMIX燃料再生组件的堆芯物理特性,设计了合理可行的30%REMIX燃料堆芯燃料管理方案。以此为基础,分析了REMIX乏燃料中钚含量和天然铀节省量变化趋势,以及REMIX燃料在燃耗状态下的源项和释热特性,并与普通AFA3G组件乏燃料进行了对比。结果表明,5次REMIX再生循环是可行的,能够显著降低天然铀消耗量,具有较好的应用价值,同时也为后续REMIX燃料加工工艺设计以及整体经济性分析提供了必要的输入条件。

深度学习引导的高通量分子筛选用于锶铯的选择性配位

试图从配位化学性质差异的角度增进对乏燃料后处理过程中锶铯分离的认识。基于对晶体结构进行数据挖掘和深度学习架构,从8种碱金属和碱土金属元素的配位结构(约3.3×10^(4)个样本)中归纳和分析锶、铯的配位化学性质,尤其是配位键长。通过引入贝叶斯优化工具,建立了高效的transformer模型,可以以很高的准确性预测配体与锶、铯离子分别的结合强度及其差异。作为概念验证,成功对配体分子结构(约9.1×10^(3)个)对锶、铯的潜在配位选择性进行排序,并为未来的配体设计确定了不同官能团对实现配位选择性的贡献度。本研究利用人工智能手段,为乏燃料后处理过程及放射化学语境中元素的配位化学信息及分离技术开发积累基础知识,为后续实验提供指导和参考。