裂隙水流-传热对高放废物处置库近场温度影响的三维离散元分析

裂隙水流-传热是高放废物处置库行为的重要影响因素。为研究裂隙水流-传热对高放废物处置库近场温度的影响,采用3DEC离散元软件计算分析了完整岩体模型和裂隙岩体水流模型对处置库近场温度分布和演变的影响。计算分析表明:由于裂隙水流的吸热降温作用,裂隙岩体模型的废物罐表面膨润土温度低于完整岩体模型的废物罐表面膨润土温度,并缩短了达到稳态所需要的时间;裂隙水流上游区域废物罐表面膨润土温度显著低于裂隙水流下游区域废物罐表面膨润土温度;在设定条件下,裂隙岩体模型的废物罐表面膨润土最高温度约为完整岩体模型废物罐表面膨润土最高温度的75%,裂隙水流速度从0.2mm/s增大到0.5mm/s,废物罐表面膨润土最高温度降低约4%。     

国内外核废物处置库近场温度场模拟预测

核废物处置后因所含的放射性核素衰变而产生的衰变热通过传导、对流以及辐射等方式从废物体向外传递,从而引起废物罐体、缓冲材料及近场围岩温度升高,导致废物体至近场围岩之间形成温度梯度。温度梯度随着时间的延续而变化,最终会影响地下水系统和核素迁移。本文对一些国家的处置库温度预测模式进行了调研,对源项、处置库模型简化、热传递数学模型和模拟结果做了初步总结,为我国拟建处置库的温度场预测提出了建议。     

对我国高放废物处置研发工作的几点建议

论述了高放废物处置库与一般地下工程设施的区别,以及处置库场址的选址工作与低中放废物处置场和核电站场址的选址工作异同点。强调要高度重视高放废物处置的安全性,这是由于高放废物毒性大、半衰期长、安全处置期长;由于处置库堆放的废物总比活度大,且高放废物处置在地下深处,因而,如果一旦处置库系统遭受破坏,就难以进行人工干预。笔者认为,在区域预选和地区(地段)预选阶段,查明场址区域地壳稳定性问题是其首要任务。文章就处置方案等若干问题进行了讨论。     

高放废物处置区域尺度岩体适宜性评价研究

岩体适宜性评价是高放废物处置库选址和设计的重要工作内容,以判断场址岩体是否满足处置库长期包容和隔离核素的功能要求。依据我国的高放废物处置概念和场址条件,提出了Q_HLW岩体适宜性评价方法,但目前Q_HLW在场址尺度展开了较为深入的研究,尚未在处置区域尺度、处置巷道及处置坑尺度建立完善系统岩体适宜性评价方法。结合芬兰地下实验室研究和处置库设计经验,建立处置区域尺度岩体适宜性评价准则Q^P_HLW,提出了裂隙带影响、地下水化学条件、岩体渗透特性、岩体强度应力比值以及岩体完整性等评价指标的取值方法,并确定岩体适宜性评价分级标准。随后,利用芬兰高放废物处置ONKALO地下实验室场址数据,测试和验证处置区域尺度岩体适宜性评价准则Q^P_HLW的合理性与可行性。最后以北山地下实验室新场场址为评价对象,开展处置区域尺度岩体适宜性评价,适宜性评价结果表明：新场场址在处置深度400～450 m及550～600 m内岩体完整性高,岩体适宜性程度高,适合布置处置巷道。     

匈牙利中低放废物最终处置库接收首批废物

【世界核新闻网站2012年12月18日报道】2012年12月5日,匈牙利在巴塔帕蒂(Bataapati)举行了国家放射性废物最终处置库的正式启动仪式。在此次仪式上,将装有9个废物桶的首个混凝土容器从位于地表的临时贮存设施转移至该处置库的第一个处置室。巴塔帕蒂处置库是由匈牙利放射性废物管理公司(Puram)历时15年时间,耗资680亿匈牙利福林(3.1亿美元)建设的     

瑞典核废物最终处置库计划得到支持

【英国《国际核工程》2004年7月刊报道】瑞典国家核燃料和废物管理公司(SKB)建设核废物最终处置库的计划得到公众支持。2个核废物最终处置库候选场址所在地区的最新民意调查表明,69%的奥斯萨玛(Osthammar)居民支持或全力支持在当地建设深层地质处置库,而奥斯卡港(Oskarshamn)的这一数字为72%。自从上一次民意调查以来,不但候选场址所在地区的公众对处置库计划的支持率一直在上升,而且周边地区的公众支持率也在增加。SKB总经理ClaesThegerstrom表示,民意调查结果表明,离候选场址越近的居民越支持处置库的建设。这可能是由于经过多年的宣…     

奥尔基卢托处置库开工建造

【英国《国际核工程》1988年6月号第5页报道】芬兰工业动力公司(TVO)已开始建造低中放废物最终处置库——ULJ处置库,该库建在奥尔基卢托核电厂厂址下基岩中。打算处置的废物包括核电厂的沥青固化块和维修产生的各种废物。装入金属桶的废物将处置到建在地表下70米的两个岩石窖中。这些窖的直径为20米,高为30米,能容纳奥尔基卢托电站30多     

芬兰庆祝中低放废物处置设施投运20周年

【英国《国际核工程》网站2012年5月9日报道】2012年5月8日是芬兰位于奥尔基洛托(Olkiluoto)的中低放废物处置库投运20周年纪念日。芬兰TVO电力公司于当天宣布,该处置库的容量目前已使用了一半。目前,这座处置库在其2个地下仓库中贮存了总计约5500m3来自奥尔基洛托核电站的中低放废物以及少量来自芬兰卫生保健部门、工业和研究部门的中低放废物。这座处置库每年接收100～180m3废物,其中2/3为低放废物,1/3为中放废物。     

英国提交德瑞格处置库扩建申请

<正>【世界核新闻网站2015年11月4日报道】英国低放废物处置库有限公司(LLW Repository)已提交在位于坎布里亚郡德瑞格(Drigg)的国家低放废物处置库新建三个处置室的规划申请。处置库公司代表核退役管理局(NDA)运营该处置库。这座处置库1959年投运,其处置的废物源自核电厂、国防机构、一般工业部门、医院和大学。1959—1995年期间,在7条处置     

高放废物处置库系统性能评价和岩石裂隙系统计算机模拟

美国桑地亚国家实验室核废物地质处置、水文地质学专家Erik Webb博士应邀于1999年3月29～30日在核工业北京地质研究院作高放废物处置库系统性能评价和岩石裂隙系统计算机模拟学术报告。学术报告的主要内容为:(1)高放废物处置库系统性能评价方法;(2)岩石裂隙系统的计算机模拟;(3)场地特性评价的决策树分析。此外,Erik Webb博士还介绍了美国桑地亚国家实验室的结构、人员组成、研究项目和管理体系以及日本核燃料循环研究所(JNC,原动燃团)高放废物地质处置研究工作的最新进展等内容。     

Analyses of disposal efficiency based on nuclear spent fuel cooling time and disposal tunnel/pit spacing for the design of a geological repository

The purpose of deep geological disposal of high-level radioactive waste (HLW) including nuclear spent fuels is to isolate and to inhibit the release of radioactive material for a long time so that its toxicity does not affect the biosphere. The main requirement for the HLW repository design is to keep the buffer temperature below 100 °C in order to maintain the integrity of the engineered barrier system. The cooling time of the spent fuels discharged from nuclear power plants is the key consideration factor for the efficiency and economic feasibility of such a repository. We analyze the spacing of the disposal tunnels and pits, the disposal area and the uranium density for the deep geological repository layout to satisfy the thermal requirement of the disposal system. To do this, thermal stability analyses of a disposal system have been performed using varying spent fuel cooling times and spacing of the disposal tunnels and pits. The results show that the time to reach the maximum temperature within the design limit of the temperature in the disposal site is likely to be shortened as the cooling time of the spent fuel becomes shorter. Also it seems that controlling the disposal pit spacing is considered more advantageous than controlling the disposal tunnel spacing to meet the allowable thermal criteria in the repository from thermal and economical points of view. The results of these analyses can be used for a deep geological repository design and detailed analyses with exact site characteristics data will reduce the uncertainty of the results.     

Efficiency analyses of the CANDU spent fuel repository using modified disposal canisters for a deep geological disposal system design

Deep geological disposal concept is considered to be the most preferable for isolating high-level radioactive waste (HLW), including nuclear spent fuels, from the biosphere in a safe manner. The purpose of deep geological disposal of HLW is to isolate radioactive waste and to inhibit its release of for a long time, so that its toxicity does not affect the human beings and the biosphere. One of the most important requirements of HLW repository design for a deep geological disposal system is to keep the buffer temperature below 100 °C in order to maintain the integrity of the engineered barrier system. In this study, a reference disposal concept for spent nuclear fuels in Korea has been reviewed, and based on this concept, efficient alternative concepts that consider modified CANDU spent fuels disposal canister, were developed. To meet the thermal requirement of the disposal system, the spacing of the disposal tunnels and that of the disposal pits for each alternative concept, were drawn following heat transfer analyses. From the result of the thermal analyses, the disposal efficiency of the alternative concepts was reviewed and the most effective concept suggested. The results of these analyses can be used for a deep geological repository design and detailed analyses, based on exact site characteristics data, will reduce the uncertainty of the results.     

Three-dimensional thermal analysis of in-floor type nuclear waste repository for a ceramic waste form

A thermal model is constructed and analyses are performed for an ‘in-floor’ type nuclear waste repository in granitic rock for a high level nuclear waste (HLW)-bearing ceramic waste form (synroc). Transient calculations for a three-dimensional (3-D) model have been carried out for both 20 and 10 wt.% HLW-bearing synroc, for surface cooling periods between reactor discharge and geological disposal varying from 5 to 40 years. This study investigates the temperature distribution in one of the boreholes of a hypothetical tunnel for a basic geometrical setting as well as the effect of varying the distance between adjacent boreholes and the distance between adjacent tunnels. The temperatures in the repository were found to be sensitive to the interim surface cooling period as well as the amount of waste loaded. The results showed that decreasing the spacing between the canisters has a more pronounced effect on the temperature field than decreasing the spacing between the tunnels.     

滑油冷却器在不同布置方式下换热特性的实验研究

以68#润滑油为工质,对整体针翅管滑油冷却器的换热和阻力特性进行实验研究。在冷却水入口温度24℃、体积流量20m³/h以及滑油入口温度55℃、体积流量6~24m³/h的实验条件下,获得了滑油冷却器竖直和水平两种布置方式时的换热特性数据,在此基础上简要分析了不同布置方式对换热特性的影响因素。实验结果表明,换热器的不同布置方式对换热特性影响较小,可根据工程使用空间选择竖直或水平布置。     

非能动余热排出热交换器流动和传热数值模拟

非能动余热排除系统(Passive Residual Heat Removal system,PRHR)是非能动核电厂的重要安全设施,在全厂断电事故下,大部分的堆芯衰变热是通过PRHR热交换器传递至内置换料水箱(In-containment Refueling Water Storage Tank,IRWST)。但PRHR热交换器属于大型非稳态换热器,其传热机理十分复杂。基于PRHR系统的重要性和复杂性,有必要研究PRHR系统的流动和传热特性。利用计算流体动力学(Computational Fluid Dynamics,CFD)软件针对非能动堆芯冷却系统试验装置中的PRHR系统进行建模计算,分析了PRHR热交换器及IRWST的流动和传热特性,发现IRWST内部沿垂直高度上呈现明显的温度分层现象,温度沿水平方向的分布趋于均匀;IRWST内部的流动主要是沿着C型传热管竖直段向上流动,流速逐渐增大,但在两相阶段,水箱上部区域流动明显增强;C型传热管上部水平段和竖直段上部区域的换热系数要明显高于其它区域,且在上部水平段与竖直段连接弯管处换热系数最大,在两相阶段,上部区域的换热系数明显增大。     

Integrated decay heat load method to analyze repository capacity impact of a fuel cycle

Assessing the needs for repository capacity from nuclear waste disposal is essential for fuel cycle development or repository development planning. As the repository capacity is mainly constrained by thermal design limits on the repository rocks, a detailed mountain-scale heat transfer calculation is needed for repository capacity impact analysis. In this paper, a simplified repository capacity impact analysis method is proposed as an alternative to performing repository scale heat transfer analysis. The method is based on the use of integrated decay heat load (IDHL) limits. The derived integrated decay heat loads were found to appropriately represent the drift wall temperature limit (200 °C) and the midway between adjacent drifts temperature limit (96 °C) under the high temperature operating mode as long as the wastes are uniformly loaded into the repository. Results indicated that the long-term integrated decay heat load (IDHL_L) and the short-term integrated decay heat load (IDHL_S) can be effectively used to represent the repository capacity impact for SNFs and HLWs, respectively. Comparisons indicated good agreement between the proposed IDHL method and the repository heat transfer analysis-based approach.     

Preliminary conceptual design of a geological disposal system for high-level wastes from the pyroprocessing of PWR spent fuels

The inventories of spent fuels are linearly dependent on the production of electricity generated by nuclear energy. Pyroprocessing of PWR spent fuels is one of promising technologies which can reduce the volume of spent fuels remarkably. The properties of high-level wastes from the pyroprocessing are totally different from those of spent fuels. A geological disposal system is proposed for the high-level wastes from pyroprocessing of spent fuels. The amount and characteristics of high-level wastes are analyzed based on the material balance of pyroprocessing. Around 665 kg of monazite ceramic wastes are expected from the pyroprocessing of 10 MtU of PWR spent fuels. Decay heat from monazite ceramic wastes is calculated using the ORIGEN-ARP program. Disposal modules consisting of storage cans, overpacks, and a deposition hole or a disposal tunnel are proposed. Four kinds of deposition methods are proposed. Thermal design is carried out with ABAQUS program and geological data obtained from the KAERI Underground Research Tunnel. Through the thermal analysis, the spacing between the disposal modules is determined for the peak temperature in buffer not to exceed 100 °C. Thermal analysis shows that the optimum spacing between the vertical deposition holes with 4 overpacks is 8 m when the disposal tunnel spacing is 40 m and optimum spacing of 2 m for horizontal disposal tunnel with 25 m tunnel spacing. Also, the spacing reduces to 6 m for vertical deposition when the double-layered buffer is used, which reduces the disposal area to one-sixty fifth (1/65th) compared with the direct disposal of spent fuels. Finally, the effect of cooling time on the disposal area is illustrated.     

矩形窄缝通道内水稳态和瞬态流动换热特性实验

以去离子水为工质,在压力0.5～5.0 MPa的范围内,对矩形窄缝通道内水稳态及瞬态流动换热特性进行了实验研究。结果表明:矩形窄缝通道在水平和竖直放置以及稳态和瞬态条件下,水的流动换热特性呈现出基本相同的规律。层流向紊流过渡区域的雷诺数(Re)为900Re1300,比常规通道提前,单相摩擦阻力系数比常规通道大;采用Dittus-Boelter公式的形式拟合得到了新的换热实验关联式,其系数较Dittus-Boelter公式的系数约小11.3%。在稳态条件下,紊流区换热系数随质量流速的增加而增大,增大趋势比较明显;换热系数随热流密度的变化不明显;压力对单相强迫对流换热特性基本没有影响。     

High burn-up operation and MOX burning in LWR; Effects of burn-up and extended cooling period of spent fuel on vitrification and disposal

Looking ahead to final disposal of high-level radioactive waste arising from further utilization of nuclear energy, the effects of high burn-up of light-water reactors (LWR) with UO₂ and MOX fuel and extended cooling period of spent fuel on waste management and disposal were discussed. It was assumed that the waste loading of waste glass is restricted by three factors: heat generation rate, MoO₃ content, and platinum group metal content. As a result of evaluation for effects of extended cooling period, the waste loading of waste glass from both UO₂ and MOX spent fuel could be increased in the current vitrification technology. For the storage of waste glass from MOX spent fuel with higher waste loading, however, those waste glass require long storage period prior to geological disposal because decay heat of ²⁴¹Am contributes significantly. Therefore, the evaluation of effects of Am separation on the storage period was performed. Furthermore, heat transfer calculation was carried out in order to evaluate the temperature of buffer material in a geological repository. The results showed, 70 to 90% of Am separation is sufficiently effective in terms of thermal feasibility of a repository.