核电厂乏燃料湿法与干法贮存技术对比研究及相关对策建议

分析了国内外主要核电国家乏燃料离堆贮存的概况,研究了乏燃料湿法贮存和干法贮存的主要特点,对乏燃料湿法贮存和干法贮存从技术特点、安全可靠性、工程经验、建设工期、电厂接驳、运行维护、经济性等方面进行了综合对比,分析两者的优劣。结合我国乏燃料的离堆贮存现状,对未来乏燃料离堆贮存技术路线提出思考和探索。     

国外最新消息

正01俄罗斯第一台VVER-1200核电机组开始商业运行3月2日,俄罗斯第一台VVER-1200核电机组,即新沃罗涅日第二核电厂1号机组正式开始商业运行。(摘编自世界核新闻网2017年3月2日报道)02英国首次使用干法乏燃料贮存技术3月13日,英国塞兹韦尔B核电站首个装有压水堆乏燃料的桶放到其干法乏燃料贮存库内。这一里程碑标志着英国首次使用干法乏燃料贮存技术。     

压水堆乏燃料中间贮存技术研究

本文通过调研主要核电国家的压水堆核电站乏燃料中间贮存与处理现状,分析研究近年来在压水堆核电站乏燃料中间贮存方面的趋势,明确乏燃料干式贮存技术将是未来压水堆核电站乏燃料中间贮存的主流。结合我国压水堆核电站乏燃料的现状并对未来核电站乏燃料贮存与处理工作提出建议。     

乏燃料干法贮存技术在英国首次得到实际应用

正【世界核新闻网站2017年3月23日报道】首个装有塞兹韦尔B机组乏燃料的容器2017年3月13日被放置在位于该电厂厂区的一座新建干法贮存设施中。这意味着乏燃料干法贮存技术在英国得到首次实际应用。塞兹韦尔B是英国唯一的压水堆机组,装机容量为1198 MWe,1995年首次并网发电。     

我国乏燃料离堆贮存需求分析

随着我国核电的大力发展,产生了大量的乏燃料。若不能妥善进行处理,会给核电发展带来不利影响。我国后处理技术的发展现状暂时无法有效缓解乏燃料大量累积造成的困境。本文按照我国的核电发展规划,结合现有的乏燃料贮存能力,计算得出了乏燃料的年产生量、累积量,以及离堆贮存需求。建议我国尽快开展压水堆乏燃料离堆贮存设施的研究工作,确保核电的安全发展。     

俄罗斯将在西伯利亚中部新建一座乏燃料贮存设施

【俄罗斯新闻信息社莫斯科6月22日电】俄罗斯一位高级官员2005年6月22日表示,俄罗斯有关部门已颁发了在西伯利亚中部城市克拉斯诺亚尔斯克(Krasnoyarsk)附近建设一座干法乏燃料贮存设施的许可证。俄罗斯联邦环境、技术与核管理监督局局长Andrei Malyshev在新闻发布会上表示,1986年已在克拉斯诺亚尔斯克采矿化学工厂建造了一座乏燃料贮存水池。与贮存水池相比,新的干法贮存设施将能对乏燃料进行更好的贮存。由于耗电量降低,因此干法贮存设施更易维护;此外,干法贮存设施还能减少二次放射性废物,并具有建设工期短、造价相对较低的特点。Malys…     

基于SCALE的压水堆乏燃料贮存水池临界安全分析

基于国际先进的核设计与安全分析计算程序SCALE,针对我国自主研发的先进压水堆乏燃料贮存水池,建立恰当的计算模型,并选取合理的保守假设,计算乏燃料水池正常贮存及事故工况下的反应性,评估计算模型的临界安全,分析该程序对我国先进反应堆乏池计算的适用性。计算结果表明该先进压水堆乏燃料贮存水池正常贮存工况及事故工况的有效增值因子均小于0.95,处于次临界状态。该设计模型可确保燃料堆内贮存区域临界状态安全可控。SCALE计算程序适用于我国自主研发的先进压水堆乏燃料水池临界安全计算。     

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。      

关于对核电厂内乏燃料干法贮存系统核安全监管要求的研究

近年来,我国一些运行核电厂乏燃料池中的乏燃料贮量已接近或达到贮存限量,这些核电厂准备增设乏燃料干法贮存项目。由于我国在该方面的核安全法规体系尚不完备,且可参考的范例较少,亟需对相关的核安全监管方式、方法进行探讨和明确。本文结合我国核安全监管要求和美国相关监管工作经验,综合考虑乏燃料干法贮存项目的可实施性和审评工作的可操作性,针对干法贮存系统在设计、运行和审批过程中必须明确的关键问题开展研究,并提出了一些可供参考的监管建议。     

基于SCALE的RFA改进型燃料组件堆内贮存临界安全分析

RFA改进型燃料组件是西屋公司设计的能应用于大功率先进压水堆的改进型燃料组件。SCALE计算程序是一款在国际上得到广泛认可的综合性建模及模拟程序包,可用于核设计与核安全分析。基于SCALE计算程序,针对大功率先进压水堆的乏燃料贮存水池,建立恰当的计算模型,并选取合理的保守假设,分析乏燃料水池正常贮存及事故工况下的临界安全。计算结果表明一区正常贮存工况keff值为0.901 29,组件跌落事故工况下,有效增值因子为0.907 93。二区正常贮存工况下,计算模型keff值为0.909 98,新燃料组件误插入事故工况keff值为0.924 07。先进压水堆乏燃料贮存水池正常贮存工况及事故工况的有效增值因子均小于0.95,处于次临界状态。该设计模型可确保燃料堆内贮存区域临界状态安全可控。     

Operational experience of on-site transport of spent nuclear fuel in Korea

Abstract

With the rapid development of the nuclear power programme in Korea, the amount of accumulated spent nuclear fuel has inevitably increased year by year. The spent nuclear fuel is being stored in on-site storage pools at the nuclear power plants. As the current storage capacity for spent nuclear fuel is insufficient, at-reactor storage is being expanded at each site with regard to optimisation of technical and economic factors. On-site transport between neighbouring reactors has been necessary to secure sufficient storage capacity for pressurised water reactor spent nuclear fuel assemblies. A complete on-site transport system has been developed, and so far more than 800 spent nuclear fuel assemblies have been transported using two kinds of transport cask.     

Characteristics of Several Equilibrium Fuel Cycles of PWR

This paper evaluated the influence of neutron spectrum on characteristics of several equilibrium fuel cycles of pressurized water reactor (PWR). In this study, five kinds of fuel cycles were investigated. Required uranium enrichment, required natural uranium amount, and toxicity of heavy metals (HMs) in spent fuel were presented for comparison. The results showed that the enrichment and the required amount of natural uranium decrease significantly with increasing number of confined heavy nuclides when uranium is discharged from the reactor. On the other hand, when uranium is totally confined, the enrichment becomes extremely high. The confinement of plutonium and minor actinides (MA) seems effective in reducing radio-toxicity of discharged wastes. By confining all heavy nuclides except uranium those three characteristics could be reduced considerably. For this fuel cycle the toxicity of HMs in spent fuel become nearly equal to or less than that of loaded uranium.     

Development of new transportable storage casks for interim dry storage

Abstract

We have started a programme to design a new type of transportable storage cask (Hitz casks) for both boiling water reactor (BWR) and pressurised water reactor (PWR) fuels for use in the new interim dry spent fuel storage project in Japan. The basic policy of this development is to use proven technology to realize a safe and cost-effective design with a high transport and storage capacity and a low fabrication cost. Since it is not permissible to change the lid gaskets at the storage facility, the double-lid system is designed to be able to use double metallic gaskets as the containment boundary for transport after the storage period; this is one of the new design features used in the casks. With the basket design we tried to achieve a capacity of 69 fuel assemblies for BWR fuel and 26 fuel assemblies for PWR fuel. Further details about these and other topics are discussed.     

Core characteristics of fast reactor cycle with simple dry pyrochemical processing

Fast reactor core concept and core nuclear characteristics are studied for the application of the simple dry pyrochemical processing for fast reactor mixed oxide spent fuels, that is, the Compound Process Fuel Cycle, large FR core with half of loaded fuels are recycled by the simple dry pyrochemical processing. Results of the core nuclear analyses show that it is possible to recycle FR spent fuel once and to have 1.01 of breeding ratio without radial blanket region. The comparison is made among three kinds of recycle fuels, LWR UO₂ spent fuel, LWR MOX spent fuel, and FR spent fuel. The recycle fuels reach an equilibrium state after recycles regardless of their starting heavy metal compositions, and the recycled FR fuel has the lowest radio-activity and the same level of heat generation among the recycle fuels. Therefore, the compound process fuel cycle has flexibility to recycle both LWR spent fuel and FR spent fuel. The concept has a possibility of enhancement of nuclear non-proliferation and process simplification of fuel cycle.     

Cost comparisons of wet and dry interim storage facilities for PWR spent nuclear fuel in Korea

As a part of an effort to determine the ideal storage solution for pressurized water reactor (PWR) spent nuclear fuel, a cost assessment was performed to better quantify the competitiveness of several storage types. Several storage solutions were chosen for comparison, including three dry storage concepts and a wet storage concept. The net present value (NPV) and the levelized unit cost (LUC) of each solution were calculated, taking into consideration established scenarios and facility size. Wet storage was calculated to be the most expensive solution for a 1700 MTU facility, and metal cask storage marked the highest cost for a 5000 MTU facility. Sensitivity analyses on discount rate, metal cask price, operation and maintenance cost, and facility size revealed that the system price is the most decisive factor affecting competitiveness among the storage types.     

轻水堆燃料关联核素的特征比值与乏燃料堆型的辨别

应用燃耗分析程序MCCOOR计算压水堆和沸水堆的栅元模型满功率运行时2种不同初始燃料富集度情形下不同燃耗深度的燃料核素成分，分析轻水堆燃料的关联核素比值与燃耗深度的关系，获得了轻水堆燃料的关联核素特征比值，并探索了由乏燃料相应核素特征比值确定其堆型的可行性。      

Status of the small modular fluidized bed light water nuclear reactor concept

The state of the art of a small modular reactor concept with a suspended core is presented. The reactor design is based on a fluidized bed concept and utilizes pressurized water reactor technology. The fuel is automatically removed from the reactor by gravity under any accident conditions. The reactor demonstrates the characteristics of inherent safety and passive cooling. Here two options for modification to the original design are proposed to increase the stability and thermal efficiency of the reactor. A modified version of the reactor involves the choice of supercritical steam as the coolant to produce a plant thermal efficiency of about 40%. Another option is to modify the shape of the reactor core to produce a non-fluctuating bed and, consequently, guarantee the dynamic stability of the reactor. The mixing of tantalum in the fuel is also proposed as an additional inhibition to the power excursion. The spent fuel pellets may not be considered nuclear waste, since they are of a shape and size that can easily be used as a source of radiation for food irradiation and industrial applications. The reactor can easily operate as a plutonium burner or can operate with a thorium fuel cycle.     

乏燃料操作大厅的剂量场随屏蔽水位的变化研究

在乏燃料水池完全丧失冷却能力和补水的事故工况下,压水堆核电厂乏燃料操作大厅内的剂量率将随着乏燃料水池水位的降低逐渐升高。本文以一典型压水堆核电厂的乏燃料水池为研究对象,采用QAD-CGGP程序,计算并分析了乏燃料操作大厅内的剂量场分布及其随水位的变化规律。计算结果表明:(1)在3.786～7.736 m水层厚度范围内,操作平台处的剂量率随水层厚度的变化不明显;(2)乏燃料水池上方的剂量率峰值位于高密格架区域上方;(3)在3.436～4.736 m水层厚度范围内,乏燃料水池上方的剂量率峰值在0.914～288 μSv/h范围内变化,并随着屏蔽水层厚度的减小呈指数递增趋势,且操作平台处剂量点的剂量率均满足乏燃料操作大厅辐射分区要求;(4)满足乏燃料操作大厅辐射分区要求所需的最低水位为+15.77 m。     

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.     

CAP1400双乏燃料贮存格架地震条件下流固耦合试验研究

CAP1400是中国自主研发的新型反应堆,其乏燃料贮存格架自由地放置在乏燃料水池内,水可在格架内部通道中流通,同时水也存在于格架与格架、格架与池壁之间的间隙中。在地震情况下,格架与流体之间的流固耦合特性非线性度非常高,很难用软件准确地模拟计算。为准确地探究CAP1400乏燃料贮存格架的流固耦合特性,建立了乏燃料贮存格架和水箱缩比试验模型,开展了振动台试验。在扫频试验中,测量了格架支脚与水箱底板间的摩擦系数;同时研究了正弦波、地震波工况下,格架与格架间的间隙对格架滑移现象、流体压力的影响。通过分析试验结果,得到了格架支脚与水池间的摩擦系数;且格架间隙越大,流体压力越小,但格架间隙对滑移现象影响不明显。本试验的结果可为CAP1400乏燃料贮存格架的流固耦合参数选取提供依据。