Numerical simulation of coupling heat transfer and thermal stress for spent fuel dry storage cask with different power distribution and tilt angles

Dry storage containers must be secured and reliable during long-term storage,and the effect of decay heat released from the internal spent fuel on the cask has become an important research topic.In this paper,a 3D computational fluid dynamics model is presented,and the accuracy of the calculation is verified,with computational errors of less than 6.2%.The thermal stress of the dry storage cask was estimated by coupling it with a transient temperature field.The total power remained constant and adjusting the power ratio of the inner and outer zones had a small effect on the stress results,with a maximum equivalent stress of approximately 5.2 kPa,which occurred at the lower edge of the shell.In the case of tilt,the temperature gradient varied in a wavy distribution,and the wave crest moved from right to left.Altering the tilt angle affects the air distribution in the annular gap,leading to the shell temperature being transformed,with a maximum equivalent stress of 202 MPa at the bottom of the shell.However,the equivalent stress in both cases was less than the yield stress(205 MPa).

Existing Condition Analysis of Dry Spent Fuel Storage Technology

As in some domestic nuclear power plants,spent fuel pools near capacity,away-from-reactor type storage should be arranged to transfer spent fuel before the pool capacity is full and the plants can operate in safety.This study compares the features of wet and dry storage technology,analyzes the actualities of foreign dry storage facilities and then introduces structural characteristics of some foreign dry storage cask.Finally,a glance will be cast on the failure of away-from-reactor storage facilities of Pressurized Water Reacto(rPWR)to meet the need of spent-fuel storage.Therefore,this study believes dry storage will be a feasible solution to the problem.

Thermal-hydraulic design and transient analysis of passive cooling system for CPR1000 spent fuel storage pool

This paper proposes a design of passive cooling system for CPR1000 spent fuel pool(SFP). Our design can effectively manage the SFP temperature not to exceed80 C. Then the transient analysis of the CPR1000 SFP with designed passive cooling system is carried out in station blackout(SBO) accident by the best-estimate thermal-hydraulic system code RELAP5. The simulation results show that to maintain the temperature of CPR1000 SFP under 80 C, the numbers of the SFP and air cooling heat exchangers tubes are 6627 and 19 086, respectively.The height difference between the bottom of the air cooling heat exchanger and the top of the SFP heat exchanger is3.8 m. The number of SFP heat exchanger tubes decreases as the height difference increases, while the number of the air cooling heat exchanger tubes increases. The transient analysis results show that after the SBO accident, a stable natural cooling circulation is established. The surface temperature of CPR1000 SFP increases continually until 80 C, which indicates that the design of the passive air cooling system for CPR1000 SFP is capable of removing the decay heat to maintain the temperature of the SFP around 80 C after losing the heat sink.

Thermodynamic Assessment of UO₂Pellet Oxidation in Mixture Atmospheres under Spent Fuel Pool Accident

For an analysis of the oxidation behavior of UO2 nuclear fuel pellet under a loss of water coolant accident in a spent nuclear fuel pool of an LWR, thermodynamic assessments of UO2 oxidation were carried out under various atmospheric conditions. In a steam atmosphere, it was assessed that UO2 would not be fully oxidized into U3O8 due to the relatively lower oxygen partial pressure, while UO2 will be fully oxidized into U3O8 in an air atmosphere. In an air and steam mixture atmosphere, the UO2 oxidation was dominantly affected by the air volumetric fraction, because of the relatively higher oxygen partial pressure of air. In addition, the effect of H2 volumetric fraction on the oxygen partial pressure under a mixture atmosphere was calculated, and it was revealed that UO2 pellet oxidation could be reduced above the critical value of H2 volumetric fraction.

Physical modeling of spent-nuclearfuel container

A new physical simulation model was developed to simulate the casting process of the ductile iron heavy section spent-nuclear-fuel container.In this physical simulation model,a heating unit with DR24 Fe-Cr-Al heating wires was used to compensate the heat loss across the non-natural surfaces of the sample,and a precise and reliable casting temperature controlling/monitoring system was employed to ensure the thermal behavior of the simulated casting to be similar to the actual casting.Also,a mould system was designed,in which changeable mould materials can be used for both the outside and inside moulds for different applications.The casting test was carried out with the designed mould and the cooling curves of central and edge points at different isothermal planes of the casting were obtained.Results show that for most isothermal planes,the temperature control system can keep the temperature differences within 6 ℃ between the edge points and the corresponding center points,indicating that this new physical simulation model has high simulation accuracy,and the mould developed can be used for optimization of casting parameters of spent-nuclear-fuel container,such as composition of ductile iron,the pouring temperature,the selection of mould material and design of cooling system.In addition,to maintain the spheroidalization of the ductile iron,the force-chilling should be used for the current physical simulation to ensure the solidification of casting in less than 2 h.

The active commissioning process for a power reactor spent fuel reprocessing pilot plant in China

The process of a power reactor spent fuel reprocessing pilot plant (hereinafter referred to as the "pilot plant") had been completed through active commissioning. Operational and technological parameters, such as shearing, dissolution, feed clarification, co-decontamination cycle, uranium and plutonium purification cycle, and the uranium and plutonium finishing facility, were identified. In addition, technical devices including extraction and mechanical equipment, electrical installation as well as instrumentation, and auxiliary systems for safety and adaptability were also verified. The commissioning results indicated that the recovery rate and decontamination coefficients of each system satisfied the designed index requirements and the qualified productions, i.e. uranium trioxide and plutonium dioxide, were produced. Monitored values at various monitoring points in the radiological protection system were within the control range and the discharge of waste water and waste gas complied with the relevant standards. This shows that independent and innovative technology for power reactor spent fuel reprocessing had been developed by our country.

Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR/CANDU synergism

High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (～22.5%), increase the energy output (～41%), decrease the quantity of spent fuels to be disposed (～2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.

Corrosion assessment for spent nuclear fuel disposal in crystalline rock,using variant cases of hydrogeological modeling

This paper presents a corrosion assessment of copper spent nuclear fuel disposal canisters in crystalline rock,using hydrogeological modeling.A simplified approach is considered,to avoid complex and time-consuming computer simulations.This simplified case is presented as a base case,with changes in the hydrogeological parameters presented as variant cases.The results show that in Taiwan’s base case,decreasing the hydraulic conductivity of the rock or decreasing the hydraulic conductivity of dikes results in a shorter transport path for sulfide and an increase in corrosion depth.However,the estimated canister failure time is still over one million years in the variant cases.

Seismic considerations for spent nuclear fuel storage in dry casks

To aid the United States Nuclear Regulatory Commission,Sandia National Laboratories (SNL) was contracted to investigate the seismic behavior of typical dry cask storage systems. Parametric evaluations characterized the sensitivity of calculated cask response characteristics to input parameters. The parametric evaluation investigated two generic cask designs (cylindrical and rectangular),three different foundation types (soft soil,hard soil,and rock),and three different casks to pad coefficients of friction (0.2,0.55,0.8) for earthquakes with peak ground accelerations of 0.25g,0.6g,1.0g and 1.25g. A total of 1 165 analyses were completed,with regression analyses being performed on the resulting cask response data to determine relationships relating the mean (16 % and 84 % confidence intervals on the mean) to peak ground acceleration,peak ground velocity,and pseudo-spectral acceleration at 1 Hz and 5 % damping. In general,the cylindrical casks experienced significantly larger responses in comparison to the rectangular cask. The cylindrical cask experienced larger top of cask displacements,larger cask rotations (rocking),and more occurrences of cask toppling (the rectangular cask never toppled over). The cylindrical cask was also susceptible to rolling once rocking had been initiated,a behavior not observed in the rectangular cask. Cask response was not overly sensitive to foundation type,but was significantly dependent on the response spectrum employed.

Challenges in spent nuclear fuel final disposal:conceptual design models

<正>The disposal of spent nuclear fuel is a long-standing issue in nuclear technology.Mainly,UO_2 and metallic U arc used as a fuel in nuclear reactors.Spent nuclear fuel contains fission products and transuranium elements,which would remain radioactive for 10~4 to 10~8 years.In this brief communication,essential concepts and engineering elements related to high-level nuclear waste disposal are described.Conceptual design models are described and discussed considering the long-time scale activity of spent nuclear fuel or high level waste.Notions of physical and chemical barriers to contain nuclear waste are highiightened.Concerns regarding integrity,self-irradiation induced decomposition and thermal effects of decay heat on the spent nuclear fuel are also discussed.The question of retrievability of spent nuclear fuel after disposal is considered.

Base-transesterification process for biodiesel fuel production from spent frying oils

The concept of converting recycled oils to clean biodiesel aims towards reducing the amount of waste oils to be treated and lowering the cost of biodiesel production. Samples of waste oils were prepared from Spent Frying oil collected from local hotels and restaurants in Khartoum, Sudan. Selected methods to achieve maximum yield of biodiesel using the waste feedstock were presented and compared. Some properties of the feedstock, such as free fatty acid content and moisture content, were measured and evaluated. Biodiesel yield recovery obtained, from Base-transesterification process about 92%. Produced Biodiesel specifications were also analyzed and discussed in Base-transesterification process. Kinematic viscosity of biodiesel was found to be 5.51 mm2·s?1 at 40?C, the flash point was 174.2?C and Cetane No of 48.19. Biodiesel was characterized by its physical and fuel properties according to ASTM and DIN V 51606 standards.

Safe Controlled Storage of SVBR-100 Spent Nuclear Fuel in the Extended-Range Future

Experience of operating reactor facilities (RF) with lead-bismuth coolant (LBC) has revealed that it is possible to perform safe refueling in short terms if the whole core is replaced and a kit of the special refueling equipment is used. However, comparing with RFs of nuclear submarines (NS), in which at the moment of performance of refueling the residual heat release is small, at RF SVBR-100 in a month after the reactor has been shut down, at the moment of performance of refueling the residual heat release is about 500 kW. Therefore, it is required to place the spent removable unit (SRU) with spent fuel subassemblies (SFSA) into the temporal storage tank (TST) filled with liquid LBC, in which the conditions for coolant natural circulation (NC) and heat removal via the tank vessel to the water cooling system are provided. After the residual heat release has been lowered to the level allowing transportation of the TST with SRU in the transporting-package container (TPC), it is proposed to consider a variant of TPCs transportation to the special site. On that site after the SRU has been reloaded into the long storage tank (LST) filled with quickly solidifying liquid lead, the TPCs can be stored during the necessary period. Thus, the controlled storage of LSTs is realized during several decades untill the time when SNF reprocessing and NFC closing are becoming economically expedient. On that storage, the four safety barriers are formed on the way of the release of radioactive products into the environment, namely: fuel matrix, fuel element cladding, solid lead and steel casing of the LST.

RY-I乏燃料运输容器维保实施及经验反馈

随着我国核工业几十年的发展,核电站及其他研究堆产生的放射性废物逐渐增多,对于放射性废物与乏燃料的运输需求也与日俱增,而乏燃料安全运输的关键在于运输容器。RY-I乏燃料运输容器是49-2反应堆和49-3反应堆的乏燃料运输容器,它由核工业总公司第二研究院设计,1990年10—12月在523厂核容器试验场对RY-I乏燃料运输容器按《放射性安全运输规定》进行了正常运输条件下和事故状态下的实验,获得了成功。主要介绍RY-I乏燃料运输容器的维保,对RY-I乏燃料运输容器进行外观检查、结构检查、密封检查,确保其在维保后能够正常使用。对维保过程中的工作进行总结,例如:维保过程中需要更换O形圈,清理容器内壁,打压做密封检查。对维保过程中出现的问题进行原因分析,改进和经验反馈。

非能动补水对AP1000乏燃料池局部硼稀释的影响分析

AP1000非能动电厂在事故后采用非能动补水的方式为乏燃料池提供冷却,同时临界安全分析要求乏池硼浓度高于最低允许硼浓度,以保证足够的次临界裕度。基于AP1000乏燃料池设计,采用CFD流体分析软件,使用多孔介质模型和组分输运模型模拟非硼化水源补水对燃料贮存区域的非均匀硼稀释影响。分析结果表明,燃料贮存区的局部硼浓度受到非硼化水源的稀释作用有所降低。在有效控制补水的前提下,局部硼浓度最低降至约2 140 ppm,高于允许的最低硼浓度值,不影响乏燃料临界安全。

基于D-D中子源的乏燃料组件钚含量测量装置设计

为了满足国际社会上对于乏燃料组件内特种可裂变材料钚的保障监督需求,以持续提供核材料信息,研究乏燃料组件内钚含量测量的非破坏性分析技术显得尤为重要。本文采用主动法,以压水堆乏燃料组件作为测量对象,D-D中子作为质询中子源,开展了乏燃料组件钚含量测量装置的设计研究。采用MCNPX软件,基于最大化探测器计数率和使各探测器计数率尽量一致的目的,对测量装置中的中子管与探测器组件距离、中子管慢化材料及其厚度、探测器组件与中子管高度差、探测器组件中慢化体厚度等关键参数进行了模拟计算。此研究为中子质询乏燃料组件钚含量测量技术研究及实验验证打下了基础。

干法后处理熔盐电解精炼过程数学模型研究

熔盐电解精炼是乏燃料干法后处理的核心工艺单元,通过数学模型探索高温熔盐电解精炼过程的化学与电化学变化,可为电解精炼工艺优化和设备设计提供参考依据。本文基于电化学热力学及物质传递公式建立了乏燃料熔盐电解精炼过程的数学模型,以铀钚锆三元合金燃料为研究对象,计算了燃料中关键元素的电极电势、分电流及物料分布随时间的变化。采用向后差分法对物料分布变化方程进行离散,通过文献实验数据对建立的数学模型进行了准确性验证。结果表明,模拟计算所得阴极沉积铀产品与实验数据的相对误差为2.80%,所建数学模型具有较好的拟合性。同时采用所建模型模拟计算了电流强度对乏燃料电解精炼过程的影响,结果表明电解速率与电流强度呈正比,不改变钚铀锆的溶解和沉积顺序。

国外乏燃料干法后处理设施进展

干法后处理技术具有介质耐辐照、临界风险低、工艺流程短、废物量小等特点,是核燃料后处理领域中适应性更高、处理对象更广的一种分离技术。干法后处理设施是实现干法后处理技术开发、验证和应用的关键场所。本文调研总结了国外干法后处理技术研发和示范设施进展,从设施建设背景、工艺基准流程、主要技术参数、设施布局设计和应用情况等多方面进行了分析和比较,并结合我国干法后处理技术发展现状和设想,提出了我国干法后处理设施发展建议。

乏燃料运输和储存容器中子屏蔽材料应用及研究现状

目的了解国内外乏燃料运输和储存容器中子屏蔽材料的类型,整理分析现有中子屏蔽材料的性能和特点,为应用于乏燃料运输和储存容器的中子屏蔽材料的研发提供一定参考。方法综述国内外应用于乏燃料运输和储存容器中子屏蔽材料的应用现状,对关键性能进行总结和比较,并提出其研究重点和发展趋势。结果目前,硼化不锈钢、碳化硼/铝复合材料、硼铝合金、聚合物基复合材料和屏蔽混凝土等中子屏蔽材料已应用于乏燃料运输和储存容器。结论随着核电厂高燃耗的发展趋势,未来乏燃料运输和储存容器对中子屏蔽材料的性能提出了更严格的要求,建议注重研发屏蔽性能优异、装配更换方便、耐辐照的中子屏蔽材料。

乏燃料后处理碱性流程的研究进展

乏燃料后处理碱性流程是用碳酸盐、氢氧化物等碱性物质的溶液作为介质进行乏燃料的溶解及铀、钚等元素的分离与纯化的方法。碱性条件下,乏燃料中的大部分裂变产物和次锕系元素并不溶解或者在溶解过程中转变为碳酸盐、氢氧化物沉淀。与已经实现工业化的PUREX(plutonium uranium redox extraction)酸性流程相比,碱性流程具有腐蚀性更小、流程更简单等潜在的优点。鉴于碱性流程的优点及其在乏燃料后处理中的潜在应用,日本、美国、俄罗斯、韩国等国家的科研人员已经围绕该流程开展了一些研究工作。本文首先介绍了各国建议的碱性流程的技术路线;然后逐一介绍了与主要工艺环节相关的基础研究的进展,包括乏燃料的氧化溶解、核素分离、试剂的回收等;最后对该领域面临的挑战和前景进行了讨论。