乏燃料后处理厂核材料衡算仿真

为优化乏燃料后处理设施的核材料衡算,寻找核材料衡算不平衡差（MUF）的主要因素,采用基于数值模拟的系统仿真方法,以核材料衡算视角构建乏燃料后处理设施核材料衡算仿真模型。改变模型工艺参数仿真不同规模的后处理设施中各环节核材料的流通量,然后以正态分布随机变量模拟各铀钚衡算测量点的随机误差,将这些带有随机特征的测量值叠加相应测量的系统误差作为核材料的仿真测量值。仿真计算结果表明,1AF中Pu、U含量测量的系统误差的方差分别占整体MUF方差的50%、40%以上,是主要误差来源。1AF的体积测量误差较小,占比MUF方差小于15%。废液中U和Pu含量很低,U和Pu含量测量的误差分别为10%和30%,对MUF方差影响不大,占比MUF方差分别小于3%和1%,废液的体积测量误差较小,占比MUF方差小于1%。U和Pu产品测量误差的方差占比MUF方差界于1AF和废液的测量之间,不是MUF误差的主要来源。     

件料核材料衡算MUF评估软件设计

以核材料衡算技术为基础,编制了件料核材料衡算MUF评估软件.该软件由数据输入模块、数据处理模块、数据查询模块、数据打印模块、系统设置模块等组成,能进行MUF值评估,根据MUF值和MUF测量方差进行t检验和置信区间估计,并能进行结果查询、打印、生成报表,同时多用户管理功能增强了信息的安全性.     

散料核设施核材料衡算与MUF评价

针对散料核设施，提供了一个核材料衡算和ＭＵＦ评价的方法。核材料衡算是以一个核材料平衡区为衡算单位，以一个闭合的核材料平衡期为限，按核材料平衡方程式（ＭＵＦ方程式）来计算。衡算的结果（即计算的ＭＵＦ值），采用概率－统计技术来进行ＭＵＦ评价，其结果可作为推断核材料是否发生了转移和大量流失的依据，并作为衡量散料核设施核材料衡算与管理性能的标志。     

高温气冷堆核材料衡算方法研究

高温气冷堆（HTR）采用球形包覆颗粒燃料元件,采用不停堆换料运行方式。因此,其运行方式、燃料元件的形式、换料方式等与压水堆核电站差别较大。HTR的特点决定了其核材料的监管方式既不同于传统压水堆,也不同于散料核设施,不易采用传统压水堆的件料管理模式和散料核设施的散料管理模式进行核材料衡算管理。为此,本文针对HTR核材料管理,提出一种适于HTR核材料衡算及其不明损失量（MUF）评价的方法。该方法根据HTR的燃料元件、运行方式和换料方式的特点,综合考虑件料和散料衡算两种模式,通过对HTR核材料衡算平衡区合理划分、关键测量点设置和实物盘存方式选取等的研究,最终选取件料+散料的衡算模式进行核材料衡算管理和评估,为HTR核材料监管提供技术基础。目前,该方法已应用于我国HTR的核材料管理,取得了预期的效果。     

乏燃料后处理厂核材料衡算与控制国际经验及关键技术启示

本文系统地调研和分析了国内乏燃料后处理厂核材料管制现状,国外商业乏燃料后处理厂核材料衡算与控制措施的实施经验和采用的关键技术,包括典型商业乏燃料后处理厂物料平衡区和实物盘存关键测量点的设置、核材料衡算与控制措施的总体设计要求、近实时衡算的概念等。根据调研结果和分析,针对我国核材料管制的现状,提出了我国在商业乏燃料后处理厂核材料管制技术准备工作的几点初步建议。     

乏燃料后处理设施核材料近实时衡算系统概念与设计

为指导后处理设施设计阶段统一部署核材料衡算、在设施运行阶段实施近实时衡算、及时反馈工艺的运行状态和趋势并探知异常情况,保障核设施核材料安全,本文在开展乏燃料后处理设施核材料衡算评价及关键技术研究的基础上,深入调研分析了核材料近实时衡算技术现状,梳理了Purex流程中核材料平衡区内过程监控的重要设备和先进仪器,以及一体化数据信息系统结构及其运行维护需求,提出我国开展乏燃料后处理近实时衡算技术研究的必要性和技术配置建议：结合传统的平衡区划分及关键测量点设置方式,以核材料重要量为目标,增补适宜的在线监控点和战略观察点,采用模型开发验证、分析方法优化评估、信息系统整合技术,在后处理设施全寿命周期内统筹管理控制Purex工艺中设备、管道、阀门、储槽中的核材料,达到近实时衡算目标。     

高温气冷堆核材料衡算方法研究

高温气冷堆(HTR)采用球形包覆颗粒燃料元件,采用不停堆换料运行方式。因此,其运行方式、燃料元件的形式、换料方式等与压水堆核电站差别较大。HTR的特点决定了其核材料的监管方式既不同于传统压水堆,也不同于散料核设施,不易采用传统压水堆的件料管理模式和散料核设施的散料管理模式进行核材料衡算管理。为此,本文针对HTR核材料管理,提出一种适于HTR核材料衡算及其不明损失量(MUF)评价的方法。该方法根据HTR的燃料元件、运行方式和换料方式的特点,综合考虑件料和散料衡算两种模式,通过对HTR核材料衡算平衡区合理划分、关键测量点设置和实物盘存方式选取等的研究,最终选取件料+散料的衡算模式进行核材料衡算管理和评估,为HTR核材料监管提供技术基础。目前,该方法已应用于我国HTR的核材料管理,取得了预期的效果。     

乏燃料后处理设施核材料近实时衡算系统概念与设计

为指导后处理设施设计阶段统一部署核材料衡算、在设施运行阶段实施近实时衡算、及时反馈工艺的运行状态和趋势并探知异常情况,保障核设施核材料安全,本文在开展乏燃料后处理设施核材料衡算评价及关键技术研究的基础上,深入调研分析了核材料近实时衡算技术现状,梳理了Purex流程中核材料平衡区内过程监控的重要设备和先进仪器,以及一体化数据信息系统结构及其运行维护需求,提出我国开展乏燃料后处理近实时衡算技术研究的必要性和技术配置建议:结合传统的平衡区划分及关键测量点设置方式,以核材料重要量为目标,增补适宜的在线监控点和战略观察点,采用模型开发验证、分析方法优化评估、信息系统整合技术,在后处理设施全寿命周期内统筹管理控制Purex工艺中设备、管道、阀门、储槽中的核材料,达到近实时衡算目标。     

滞留量测量技术研究的进展

如何测定核设施工艺设备内的滞留量一直是核材料衡算过程中的技术难题,因而直接影响核材料生产过程的闭合衡算,同时也是一个重要的不安全因素。不同的工艺过程、设备和设施环境,以及现场各种变化因素,都会使测量与分析的难度增加。 本课题组针对某工艺过程,经过现场调查,确定了     

田湾核电站（TNPP）核材料平衡区（MBA）之浅析

在核电站核材料衡算管理中，核材料平衡区的划分是根据核电站自身的结构特点。在满足电站核材料衡算管理要求的前提下人为设定的区域，旨在能够方便、准确掌握核电站平衡区内的核材料存量及变化活动。本文从田湾核电站结构的特点出发，介绍田湾核电站核材料平衡区的划分及其特点。并与相应的参考电站一巴拉科夫核电站(BNPP）进行比较说明．     

用于核保障中核材料衡算的液体闪烁体中子多重性测量装置

本工作优化设计建立了1套可用于核保障中核材料衡算的液体闪烁体中子多重性测量装置，并基于该装置开展了性能测试。结果表明，装置运行状态稳定，各项指标均在可接受范围内。同时开展了对²⁵²Cf源和标准Pu样品的实验室模拟测量等中子多重性测量研究。结果表明，该测量装置的探测效率好于15%，测量相对标准偏差为8.6%，表明在条件允许时，通过长时间信号采集，该中子多重性测量装置有能力替代基于³He管的中子多重性测量装置，通过中子多重性分析完成核保障中核材料衡算定量测量任务。     

钚质量的量热法测量

量热计是核保障领域中核材料衡算非破坏性直接测量最方便和最准确的仪器之一。本工作应用一种测量钚的热功率范围为0.1～15W的量热计测量了3个²³⁸Pu含量已知的钚样品。²³⁸Pu质量测量值与参考值相对偏差均在-1.7%以内。     

Automatic Determination of About 100 mg Uranium by Developed Potentiometric Titration Based on Conventional Method

Accurate determination of uranium is significant in the nuclear fuel production, accountancy, nuclear safeguards and other procedures of nuclear fuel cycle. Electrochemical method based on the redox titration is conventionally used for the determination of about a gram amount of uranium in different nuclear fuel materials.     

The Transport of Plutonium Nitrate Liquor in the UK

Abstract

The PFR fuel cycle reasons for plutonium nitrate liquor transport are discussed, identifying the pre 1980 ‘one litre’ bottle associated with nuclear R&D activities and then the 250 litre PUNIT transport container to support the PFR fuel cycle. The PUNIT vessel design is discussed, addressing, in particular, the need for clean filling and criticality design features together with the identification of the IAEA transport regulations criteria and the late 1970's testing programme—the full scale drop tests and the model fire tests. Radiolysis is discussed together with the need for inert gas padding. The filling facility and operation are then outlined together with the fissile material accountancy requirements. The shipments and emergency arrangements are then described drawing attention to time and weather constraints. The risk assessment is discussed, together with reference to the Competent Authority (DoT) ‘authorisation’ and any constraints. A history of the amount of material transferred is then given—relating that to the reprocessing plant operations since 1980. The most satisfactory performance over 12 years is noted.     

Development of PRIDE UNDA for safeguards application

In order to enhance the safeguardability of a pyroprocessing facility, the Korea Atomic Energy Research Institute (KAERI) has been endeavoring to develop more efficient and effective safeguards technologies for nuclear material accountancy (NMA), process monitoring, and containment and surveillance (C/S). NMA has two components: destructive analysis (DA) and non-destructive assay (NDA). Although DA is more accurate, it is typically time-consuming and cost-intensive. NDA, on the other hand, can provide reasonable accuracy on a real-time or near-real-time basis, which maximizes the utilization efficiency of a facility. In this study, the PRIDE (PyRoprocessing Integrated inactive DEmonstration) UNDA (unified non-destructive assay) was developed for testing NDA techniques at PRIDE, a demonstration facility within KAERI for integrated pyroprocessing using depleted uranium and surrogate materials. Each component of the PRIDE UNDA (i.e., neutron, gamma-ray, and mass measurement systems) was characterized and calibrated using calibration sources and standard weights as well as nuclear material used in the facility (depleted uranium). It is expected that in the near future, the PRIDE UNDA will be installed and tested with various types of process materials.     

Isotope correlation technique for Pu mass analysis of PWR UO2 spent fuel rods

A simple and fast method of nuclear material accountancy of pressurized water reactor (PWR) UO₂ spent fuel rods for safeguards application was developed utilizing the isotope correlation between the amounts of ¹³⁷Cs and total Pu. To this end, the following steps were taken: (1) as much destructive analysis (DA) data as possible for segments taken from a PWR UO₂ spent fuel rod were aggregated from publicly available data sources; (2) the DA data were corrected so as to have the same cooling time (i.e., CT = 0 y) and analyzed for outliers; (3) an equation converting the ¹³⁷Cs amount to the Pu amount was obtained by regression analysis with logarithmic curve fitting; and (4) the error in determining the Pu amount was evaluated for the imposition of a limit on the range of burnup (BU) or initial enrichment (IE). It was found that the averaged % error in calibration was determined to be 3.88% ± 2.68% (= mean ± 1 standard deviation) for the BU range over 30 GWd/tU and falling with increasing BU range. On the other hand, there was no benefit in applying the limit of the IE range. Lastly, the Pu-mass difference between various methods was compared and it was found that the difference can be incurred up to 11.4%, according to the choice of method. In conclusion, the proposed isotope correlation technique could be used for input material accountancy with reasonable uncertainty.     

Neutron kinetics of fluid–fuel systems by the quasi-static method

The quasi-static method for the neutron kinetics of nuclear reactors is generalized for application to neutron multiplying systems fueled by a fluid multiplying material, typically a mixture of fissile molten salts. The method is derived by the application of factorization formulae for both the neutron density and the delayed precursor concentrations and the projection of the balance equations upon a weighting function. A physically meaningful weight can be assumed as the solution of the adjoint model, which is constructed for the situation considered, including delayed neutrons. The quasi-static scheme is then applied to calculations of some transients for a typical configuration of a molten-salt reactor, in a multigroup diffusion model with a one-dimensional slug-flow velocity field. The physical features associated to the motion of the fissile material are highlighted.     

Distribution Equilibrium of Tritium between Adsorbed Water on Molecular Sieve 5A and Environmental Water

In order to calculate airflow around a nuclear site, a new weighted interpolation method for sparse wind data at various terrain heights is developed. This method employs weighting functions for vertical distance and the topographic barrier between the station and a grid point, in addition to a weighting function for horizontal distance which is employed in usual methods. The new weighting functions are developed from the analysis of field measurements taken in complex terrain. This method can represent the wind-fields from the bottom to the upper boundary better than usual methods which use only the horizontal weighting factor, even when the upper wind data are not available.     

多方向加权方法在D-T聚变中子能谱测量中的应用

中子能谱是研究和诊断核反应过程特性最重要的特征量之一。建立了一种新的多方向加权方法用于D-T聚变中子能谱测量:在反冲质子出射方向的多个不同角度上,同时布置探测器,最终的中子能谱由各方向所获取的反冲质子能谱结合相应的权重值确定。Geant4模拟结果显示,多方向加权方法可以提升探测效率和求解结果精度。利用多方向加权方法对高斯分布中子源以及实际的D-T聚变中子能谱进行测量模拟与求解分析,分析结果验证了该方法的可行性和有效性。