福岛乏燃料水池事故探讨

日本福岛核事故暴露出乏燃料水池安全的重要性和严峻性,乏燃料水池的安全监管应给予高度重视.本文描述了日本福岛第一核电厂乏燃料水池的基本情况,简要分析了4号机组乏燃料水池的事故起因和乏燃料源项,最后总结了从此次事故中汲取的经验教训.     

乏燃料桶外部剂量的MCNP分析

乏燃料运输容器是运输反应堆乏燃料的专用设备.文章使用MCNP4C软件计算西屋MC-10乏燃料桶的外部剂量率,方法是基于网格的定义,得到桶周围外部整体剂量率,并与前研究工作者Georgeta Radulescu的计算结果相比,结果一致性符合较好,使用网格造成机时耗费较大,也存在一定误差,但能够比较全面,详尽地得到关于乏燃料桶外部剂量的一些数据,可以对工作人员确定乏燃料桶外部剂量有一定的参考意义.     

核电站乏燃料水池遭受恐怖袭击后果评价

以典型的百万千瓦级压水堆核电站为例,介绍了乏燃料水池及乏燃料组件的特征,分析了乏燃料水池遭受恐怖袭击的情景和释放源项,并在此基础上使用后果评价程序MACCS进行了计算。结果表明在所有乏燃料组件均燃烧、仅最后卸出的一炉组件燃烧和最后卸出的三炉组件间隙释放的三种情景下,有急性死亡危险的区域半径分别约为6km、3km和0km,有效剂量超过50mSv的区域半径分别约为80km、34km和9km,隐蔽的可防止剂量超过10mSv的区域半径分别约为100km、48km和11km。     

乏燃料干法贮存容器屏蔽性能研究

乏燃料干法贮存容器的屏蔽设计是关系到乏燃料干法贮存设施辐射安全的重要影响因素。以我国自主研发的乏燃料干法贮存容器为对象，重点围绕屏蔽性能目标的合理确定、屏蔽计算方法、计算假设的选择以及计算结果的分析等多方面展开研究。计算分析表明，自主研发的的乏燃料干法贮存容器的屏蔽性能良好，满足辐射安全要求。      

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。     

蒙特卡罗方法在乏燃料运输容器屏蔽中的应用

本文主要应用蒙特卡罗方法进行了屏蔽计算的研究.在开展蒙特卡罗方法的实际应用中,对大亚湾第一炉换料乏燃料运输容器进行了屏蔽计算.源项应用ORIGEN-Ⅱ程序的计算结果,应用MCNP程序对乏燃料运输容器(STC)进行屏蔽计算,并将计算结果与实测值进行了比较分析.考虑到今后可能会运输更高燃耗的乏燃料,本文对这类乏燃料的运输也进行了计算.     

迷宫屏蔽剂量计算方法的比较研究

本文在NCRP第49、51和151号报告基础之上,概述了低能和高能加速器迷宫剂量率的理论计算方法。应用此种方法,分别对同方威视技术股份公司密云CT厂房的多折迷宫和军事医学科学院辐照实验室的弧形迷宫进行了理论计算。通过与现场测量实验和蒙特卡罗模拟计算结果的对比得出,该方法只要参数选取合理,对于一般迷宫剂量率的估算,可以节省大量计算时间,得到较为保守准确的结果。希望可以为今后迷宫的设计和剂量估算方法提供参考。     

压水堆核电厂乏燃料组件γ源强研究

本文运用ORIGEN-ARP计算研究了压水堆核电厂反应堆平衡循环的乏燃料组件的γ源强,对影响γ源强的因素,包括总燃耗、各循环燃耗比例和能群结构划分方式进行了分析。分析结果表明:乏燃料组件中,裂变产物产生的γ源强始终占主要部分。在卸料后的不同冷却时刻,γ总源强与总燃耗或末端燃耗密度存在正比关系。采用不同γ能群结构划分方式对γ总源强计算结果的影响较大。     

某型乏燃料运输容器中子辐射屏蔽性能检测

中子辐射水平测量的可靠性是辐射屏蔽性能检测的难点。本文采用便携式中子测量仪和多球谱仪对某型乏燃料运输货包外部中子辐射水平进行了测量,并基于SCALE程序计算得到的乏燃料中子源项,采用MCNP程序模拟计算得到货包外部中子辐射水平。对测量结果和计算结果进行比较,分析相关影响因素,提出了优化测量方案的建议。     

用于乏燃料组件上下端部活化源项分析的截面库研究北大核心CSCD

基于乏燃料组件上下端部活化源项精细化分析的需求,自制精细化问题相关截面库,并通过对比分析,研究不同截面库应用特点:活性区截面库和热中子截面库应用于活化源项计算时均存在较大的计算误差;组合截面库对共振群和快群截面的处理偏保守,当程序应用于上下端头活化源项计算时,分析结果偏保守;基于问题相关中子能谱自制的反应截面更加接近真实情况,基于此的计算结果也最接近真实情况。     

Prediction of temperature and water level in a spent fuel pit during loss of all AC power supplies

A prediction method for water temperature in a spent fuel pit of a pressurized water reactor (PWR) has been developed to calculate the increase in water temperature during the shutdown of cooling systems. In this study, the prediction method was extended to calculate the water level in a spent fuel pit during loss of all AC power supplies, and predicted results were compared with measured values of spent fuel pools in the Fukushima Daiichi Nuclear Power Station. The calculations gave reasonable results, but overestimated the decreasing rate of the water level and the water temperature. This indicated that decay heat was overestimated and evaporation heat transfer from the water surface was underestimated. Results of calculations with 80% decay heat and 155% (Unit 4 pool) or 230% (Unit 2 pool) evaporation heat flux were in good agreement with measured values. The data-fitted evaporation heat fluxes agreed rather well with the evaporation heat transfer correlation proposed by Fujii et al.     

乏燃料公海铁联运核应急体系建设研究

公海铁联运作为解决大宗乏燃料远距离运输的最佳方案,在国际上是一种较为普遍的运输模式,如果未来我国采用该运输模式,需探索相关核应急工作思路。本文调研梳理了国内外乏燃料公海铁联运核应急相关法规标准,参考借鉴国外乏燃料运输相关实践,提出我国乏燃料公海铁联运核应急体系建设相关工作建议。     

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.     

我国乏燃料运输现状探讨

随着我国经济的持续发展,核能作为安全、清洁能源在我国能源战略中地位日益突出。在保证安全的前提下,我国核电机组按照国家规划合理增加,乏燃料的产量也将逐步增加。根据我国核电站乏燃料贮存及外运规则,以及我国核电站主要位于东部沿海,而乏燃料后处理厂处在西北腹地这一国情,必将面临乏燃料的大量、长距离及安全运输的问题。乏燃料运输作为联接核电站与后处理厂或最终处置场的纽带,在维持核燃料循环体系的正常运行上发挥至关重要的作用。对国内外乏燃料运输涉及的运输方式、运输容器、运输安全监管及事故应急体系等问题进行了分析和讨论,对我国乏燃料运输中存在问题的解决提出了建议。     

高燃耗对乏燃料贮存和运输的影响研究

本文对高燃耗对乏燃料包壳结构完整性的影响进行了分析。探讨了影响包壳结构完整性的重要温度限值,即燃料包壳温度限值、包壳溶解温度以及韧脆转变温度(DBTT)。给出了分析包壳结构完整性的方法,对拟在干式贮存设施内贮存超过20年的容器性能及贮存后运输时乏燃料组件的结构完整性进行了分析,并给出了相关建议。     

Study on modeling of spray cooling for spent fuel pool accidents

In Japan, spray equipment is prepared in spent fuel pools (SFP) in accordance with the regulatory requirements to mitigate fuel damage in the event that the water level of SFP cannot be maintained. In order to evaluate the spray coolability of fuel assemblies in SFP accidents, the spray cooling experiments were conducted under the SFP conditions. The experimental facility contains one mock-up BWR fuel assembly with full-length 7 × 7 heater rods in a mock-up SFP rack. The measured surface temperatures indicate that the spray injection results in the top-down quench and the precursory cooling, which are consistent with the spray-cooling mechanism that has been revealed by previous studies investigating reactor core spray. Further, the numerical simulations of the experiments were conducted using the TRACE code to examine the applicability of system codes for evaluating the spray coolability of SFPs. Although the TRACE calculation with a simple analytical model reproduced the top-down quench by spray injection as observed in the experiments, some qualitative differences were found between the experiments and calculations. The causes of these differences were revealed and the applicability of system codes were discussed.     

Study on the two-way corrugated aluminum honeycomb as a filler material in impact limiters for spent fuel transport casks

Impact limiters installed on nuclear spent fuel transport casks are used to absorb energy and limit overload during transport and accidents. The two-way corrugated aluminum honeycomb, a new kind of filler material, is designed based on the action mechanism of the impact limiter. Quasi-static compression tests are used to compare the properties of the honeycomb and the traditional filler material, paulownia wood. Experimental results indicated that the compressive and energy absorption properties of paulownia wood with axial wood grain and radial wood grain were very different. Also, the moisture content of paulownia wood led to a significant decrease in its properties. The two-way corrugated aluminum honeycomb, as a new porous material, showed better compression and energy absorption properties than paulownia wood in the x, y, and z directions. The peak stress (σ_pe), platform stress (σ_pl), and energy absorption capacity (W_EA) were 2.10 times, 2.07 times, and 1.69 times higher than that of paulownia wood with axial wood grain. The two-way corrugated aluminum honeycomb is a filler material that has essential application value in impact limiters of spent fuel transport casks.     

Experimental study on elemental behaviors in fluorination of nuclear spent fuel with flame reactor

Our proposed spent nuclear fuel reprocessing technology named FLUOREX is a hybrid system based on reprocessing technologies of fluorination and solvent extraction for light water reactor fuel. In the current research, we experimentally clarified solid–gas transfer behaviours of the fluorides in the FLUOREX process and identified the volatile and non-volatile compounds in the fluorination. We carried out a fluorination experiment for simulated spent nuclear fuel and solid separation from the UF₆ gas stream. The distribution ratios of fission product elements in the experimental apparatus were evaluated. Molybdenum, Te, Nb, and Ru were volatilized by fluorination and they accompanied the UF₆ gas. However, 22.9% of the Ru and 3.4% of the Nb were retained as solids in the experimental apparatus, contrary to the fact that their partial pressures in the experiment were lower than their vapor pressures. Rubidium, Sr, Zr, Ce, and Nd were completely recovered as solid fluorides, and these results agreed with the prediction based on boiling points of their fluorides. Antimony was completely recovered as a solid; nevertheless, the boiling point of antimony pentafluoride was lower than the process temperature, and that was attributed to the formation of a non-volatile antimony oxyfluoride.