燃耗信任制下燃耗计算对临界计算的偏差及不确定度的研究

为量化燃耗信任制中燃耗计算传递给临界计算的不确定度,本文基于参数统计法对燃耗计算的核素偏差及偏差不确定度展开分析,并以蒙特卡罗（MC）抽样方法计算的kinf不确定度为基准,比较不同抽样方法对临界计算不确定度的影响。结果表明,核素偏差与偏差不确定度是随样品燃耗变化的分段函数。对于临界计算,拉丁超立方抽样（LHS）方法与MC抽样方法的kinf不确定度计算结果吻合较好,且LHS方法可考虑参数间的相关性,计算结果更真实,可进一步提升电厂的经济性。     

乏燃料贮存燃耗信任制计算的敏感元素分析

燃耗信任制的应用分析需要计算乏燃料系统的反应性,为了保证临界安全需要保证分析结果的包络性,为此需要考虑到各种不同燃耗因素对于分析结果的影响。采用SCALE程序包中的STARBUCS模块,通过对OECD/NEA发布的Phase-IA及Phase-IIA基准题的验证分析,研究了不同信任等级、堆芯运行参数等因子对乏燃料贮存系统临界安全的敏感性分析,并得出有益的结论。     

乏燃料后处理溶解过程核临界安全初步分析

通过建立合理的空间分布模型,对后处理厂乏燃料溶解不同阶段的核临界安全问题进行分析,同时对重要的核临界安全参数给予影响评价。结果显示,在仅考虑易裂变核素形态转变的理想情况下,溶解初期为最危险状态;温度升高和硝酸浓度增大对系统的影响为负效应,影响均小于4%;可溶中子毒物的加入与燃耗信任制技术的应用能大幅提高系统的经济性,影响均可达到30%。     

田湾核电站乏燃料干式贮存临界安全计算研究

以CASTOR 1000/19干式贮存容器装载田湾核电站六角形乏燃料组件为例,研究六角形乏燃料干式贮存的临界安全问题。基于新燃料假设,应用MONK9A程序对贮存容器满装载乏燃料进行不同工况下k_eff的计算。计算结果表明：正常工况下,k_eff远小于临界安全限值,是临界安全的;事故工况下,当²³⁵U富集度大于3.15%时,系统存在临界安全风险,须减少乏燃料装载量来确保临界安全。考虑燃耗信任制后,采用相同的模型计算得出贮存容器满装载的参考装载曲线,按此曲线要求装载能确保所有工况下的系统临界安全。采用燃耗信任制技术提高了贮存容器的利用率。该研究可为田湾核电站采用乏燃料干式贮存方案提供依据。     

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。     

燃耗信任制临界计算中保守性因素研究

在运用燃耗信任制技术进行乏燃料储存、运输等环节的临界安全分析时,临界计算所采用的条件是否具有足够的包络性十分关键。本文借助于OECD/NEA发布的若干燃耗信任制临界安全基准题,使用SCALE5.1软件中的STARBUCS模块进行分析,对信任核素选取、乏燃料冷却时间以及端末效应等因素对乏燃料系统临界安全性的影响进行了研究,得出了各参数保守性的有关结论。     

Active reduction of the end effect by local installation of neutron absorbers

In the burnup credit analyses of interim or long-term spent fuel (SF) storage facilities and transport casks, when the average burnup value is greater than approximately 30 GWd/t, the neutron multiplication factor becomes greater if we consider the axial burnup distribution of the spent fuel assembly rather than assuming an average burnup. This phenomenon is called the “end effect” and it is one of the main technical issues in burnup credit research. The end effect is characterized by an increase of the neutron flux around the end regions of the spent fuel assemblies in the criticality calculation. However, such increase of the neutron flux has not been observed in experiments using actual spent fuel assemblies.     

田湾核电站乏燃料水池采用燃耗信任制的计算研究

以田湾核电站(TNPS)2×5排列的贮存格架构成的乏燃料水池为例,研究采用燃耗信任制技术的密集贮存和临界安全问题。采用MONK9A程序计算分析不同富集度、不同燃耗的乏燃料装载情况下系统的k_eff. 根据系统k_eff随不同初始富集度燃料的燃耗变化情况给出了水池的参考装载曲线。采用燃耗信任制技术的密集贮存方案能提高贮存能力31%。     

An analysis of burnup reactivity credit for reactor RA spent fuel storage

The need for increasing the spent fuel storage capacity has led to the development of validated methods for assessing the reactivity effects associated with fuel burnup. This paper gives an overview of the criticality safety analysis methodology used to investigate the sensitivity of storage system reactivities to changes in fuel burnup. Results representing the validation of the methods are also discussed. As an example of the application of this methodology an analysis of the burnup reactivity credit for the three-dimensional model of the reactor RA spent fuel storage is described.     

压水堆核电厂乏燃料组件源项计算分析

核燃料贮存、运输以及后处理过程中的安全是构成核与辐射安全的重要内容,为保证安全性,提高运输经济性,减小后处理厂对环境的排放,须获得乏燃料组件的包络源项,因此,采用ORIGEN-ARP程序分析组件运行历史、初始富集度、燃耗深度等参数对源项的影响。运行历史在卸料初期对源项略有影响,可采用合适的保守因子予以包络,在冷却一定时间后,其影响可忽略不计;初始富集度、燃耗深度均不同的组件须经对比计算以获得包络源项。计算表明：在目前核电厂乏燃料组件中,²³⁵U初始富集度为4.45%、燃耗深度为55 GW•d/tU的AFA-3G型组件源项是包络的,可作为乏燃料水池、运输容器设计,以及后处理厂排放源项分析的初始源项。     

Taking burnup credit into account in criticality studies: the situation as it is now and the prospect for the future

As enrichment of the fuel has become higher than the limits used at the designing stages, it seemed necessary to consider fuel depletion during irradiation to guarantee the criticality safety for relatively high enriched fuels transportation, storage or reprocessing. This burnup credit will make it possible to use the devices for spent fuels which are initially relatively high enriched. For that purpose, a method was developed considering: (i) partial Uranium-and-Plutonium burnup credit in the criticality studies, and (ii) a conservative assumption concerning the axial profile; this actinides-only method was supported by an experimental program called HTC. The method was accepted by the French Safety Authority. Moreover, in order to reduce again the calculated values of the reactivity for irradiated fuels, a French working group was set up in 1997 to define a conservative method which enables industrial companies to take burnup credit into account with some of the fission products and using a more precise profile. The work of this group has been divided into four tasks related to: the determination of (i) the composition of the fuel, (ii) a conservative profile, (iii) a conservative irradiation history, and (iv) the calculation scheme. This work is also supported by experimental programs related to the validation of the fission products effects, in terms of reactivity.     

Calibration of Burnup Monitor Installed in Rokkasho Reprocessing Plant

Rokkasho Reprocessing Plant uses burnup credit for criticality control at the Spent Fuel Storage Facility (SFSF) and the Dissolution Facility. A burnup monitor measures nondestructively burnup value of a spent fuel assembly and guarantees the credit for burnup. For practical reasons, a standard radiation source is not used in calibration of the burnup monitor, but the burnup values of many spent fuel assemblies are measured based on operator-declared burnup values. This paper describes the concept of burnup credit, the burnup monitor, and the calibration method. It is concluded, from the results of calibration tests, that the calibration method is valid.     

加速器驱动的10 MW次临界反应堆物理方案研究

加速器驱动的次临界系统（ADS）是未来最有可能实现工业化嬗变核废料的装置。通过设计1个10 MW的ADS物理方案,研究ADS的嬗变能力。采用MCNPX和ORIGEN的耦合程序,利用基于ENDF6.8处理所得的6个温度（300、600、900、1 200、1 500、1 800 K）下连续能量核数据库,计算得到ADS随燃耗时间变化的有效增殖因数k_eff、功率峰因子和质子束流强度。同时通过计算给出了该设计方案下ADS燃料多普勒系数、冷却剂空泡系数和有效缓发中子份额,利用这些物理量研究了该ADS方案的安全特性,并通过燃耗计算研究了ADS的嬗变能力。结果表明,在1 000 d燃耗时长内,k_eff和质子流强随时间的波动较小,燃料燃耗深度较浅,系统可提升功率运行,在假想事故下系统能保持次临界状态。系统嬗变支持比约为8。     

UO2燃料燃耗对高放废物管理的影响研究

从乏燃料的不同燃耗引起放射性和化学组成的变化出发,分析乏燃料经后处理后的衰变热、Mo及贵金属含量对玻璃固化工艺和玻璃固化体储存的影响,计算得到了不同燃耗乏燃料制得的高放玻璃的数量。计算结果认为：对于冷却8 a的乏燃料,决定玻璃固化体包容量的不是高放主组分的热功率;对于燃耗小于40 GW•d/tU的乏燃料,决定玻璃固化体包容量的是Mo元素含量;当燃耗大于45 GW•d/tU时,贵金属含量成为决定玻璃固化体包容量的主要因素,同时UO₂燃料燃耗与高放玻璃固化体数量上存在线性关系,燃耗增加会导致高放废物玻璃固化体数量增加。随着燃耗的增加,以Mo含量及贵金属含量计算得到的玻璃固化体数量比以衰变热计算得到的玻璃固化体数量多,因此,高放废物玻璃固化前将Mo及贵金属进行分离有利于减少高放废物玻璃固化体数量。对于UO₂燃料,燃耗加深对于高放废物玻璃固化体暂存时间几乎无影响。     

Problems and Experience in Guaranteeing the Nuclear Safety of the Storage of Spent Fuel from Nuclear Power Plants

Nuclear-safety problems are examined and the results of investigations of nuclear safety of storage sites are presented for spent nuclear fuel from nuclear power plants. The initial events of anticipated and unanticipated accidents, methods and errors in the calculation of k _eff taking account of burnup to ensure nuclear safety, the possibility of measuring k _eff of storage sites experimentally, and new forms of fuel with a consummable absorber are calculated.     

Analyses of used BWR fuel assay data with the integrated burnup code system SWAT4.0

Criticality safety of the fuel debris from the Fukushima Daiichi Nuclear Power Plant is one of the most important issues, and the adoption of burnup credit is desired for criticality safety evaluation. To adopt the burnup credit, validation of the burnup calculation codes is required. Assay data of the used nuclear fuel irradiated by the Fukushima Daini Nuclear Power Plant Unit 2 are evaluated to validate the SWAT4.0 code for solving the BWR fuel burnup problem. The calculation results revealed that the number densities of many heavy nuclides and fission products show good agreement with the experimental data, except for those of ²³⁷Np, ²³⁸Pu, and samarium isotopes. These differences were considered to originate from inappropriate assumption of void fraction. Our results implied overestimation of the (n, γ) cross-section of ²³⁷Np in JENDL-4.0. The Calculation/Experiment – 1 (C/E–1) value did not depend on the type of fuel rod (UO₂ or UO₂–Gd₂O₃), which was similar to the case of PWR fuel. The differences in the number densities of ²³⁵U, ²³⁹Pu, ²⁴⁰Pu, ²⁴¹Pu, ¹⁴⁹Sm, and ¹⁵¹Sm have a large impact on k_eff. However, the reactivity uncertainty related to the burnup analysis was less than 3%. These results indicate that SWAT4.0 appropriately analyzes the isotopic composition of BWR fuel, and it has sufficient accuracy to be adopted in the burnup credit evaluation of fuel debris.     

未来十年的临界安全技术

半个世纪以来,临界安全技术主要借鉴核武器与反应堆技术,未来十年临界安全将面临含易裂变材料废物贮存、长期地置处置、系统临界安全分析、贫铀在临界安全中的利用,法规改进,临界实验与基准、燃耗信用的应用、乏燃料元件干法贮存,计算方法改进,次临界袷度的统一标准等问题,本文展望了未来十年的临界安全技术,并对其中一些问题作了初步探讨。     

燃耗数据库基准检验方法研究

燃耗数据库基准检验方法对于研制高准确度的燃耗数据库至关重要。本文以TAKAHAMA 3压水堆辐照后检验实验中SF95样品的建模为例，研究了建模要素对燃耗计算的影响，确定了燃耗实验建模的方法，开展了燃耗信用制研究感兴趣的锕系和裂变产物核素积存量计算值与实验值的比对。比对结果显示，主锕系核素计算偏差小于2%，大部分次锕系核素偏差小于10%，大部分重要裂变产物核素偏差小于5%。本文还对125Sb积存量随燃耗深度变化规律进行了理论分析，确认了破坏性放化实验测量结果存在缺陷，并进一步获得了125Sb积存量的修正值，使计算偏差从接近170%下降到20%以内。本次研究表明，燃耗数据库基准检验研究不仅需发展适当的燃耗实验建模方法，还需对实验数据进行适当的评价。     

Thinking outside the box: options for transport of high burnup spent fuel

Abstract

The current uncertainty surrounding the licensing and eventual opening of a long term geologic repository for the nation’s civilian and defense spent nuclear fuel and high level radioactive waste has shifted the window for the length of time spent fuel could be stored to periods of time significantly longer than the current licensing period of 40 years for dry storage. An alternative approach may be needed to the licensing of high burnup fuel for storage and transportation based on the assumption that spent fuel cladding may not always remain intact. The approach would permit spent fuel to be retrieved on a canister basis and could lessen the need for repackaging of spent fuel. This approach is being presented as a possible engineering solution to address the uncertainties and lack of data availability for cladding properties for high burnup fuel and extended storage time frames. The proposed approach does not involve relaxing current safety standards for criticality safety, containment, or permissible external dose rates.