Recycling and Transmutation of Spent Fuel as a Sustainable Option for the Nuclear Energy Development

The objective of this paper is to discuss the option of recycling and transmutation of radioactive waste against once-through fuel cycle based on uranium feed under the perspective of sustainability. A qualitative analysis was carried out to compare the fuel cycles considering different options for burning and recycling transuranic and fission products utilizing accelerator driven systems, fast reactors, and light water reactors. The results show that recycling and transmutation fuel cycles are more attractive than the current once-through fuel cycles from the point of view of sustainability. The main conclusion is that the decision about the construction of deep geological repositories for spent fuel disposal must be reevaluated.

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.

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.

乙醛肟还原Np(Ⅵ)的机理研究

在Purex流程中,调控Np的价态能实现乏燃料中镎的分离。乙醛肟(CH_(3)CHNOH)作为无盐还原剂可有效将Np(Ⅵ)还原为Np(Ⅴ),但微观还原机理尚不清楚。CH_(3)CHNOH存在顺式(Z)和反式(E)异构体,这两种异构体对Np(Ⅵ)可能具有不同的还原能力和反应过程。本研究利用标量相对论密度泛函理论分别探讨了Z/E-CH_(3)CHNOH还原Np(Ⅵ)的反应机理。反应的热力学结果表明,Z-CH_(3)CHNOH还原Np(Ⅵ)的过程比E-CH_(3)CHNOH更有利,这可能归因于前者形成更多的氢键和反应过程中结构变化较小。动力学结果表明,两种同分异构体还原Np(Ⅵ)的决速步能垒非常相近,分别为22.36、23.03 kcal/mol,表明两者的还原能力基本一致。键长分析结果表明,Z/E-CH_(3)CHNOH还原2个Np(Ⅵ)的过程都伴随着相关键的断裂与形成。第1个Np(Ⅵ)还原属于氢原子转移,第2个Np(Ⅵ)还原是水参与的单电子转移。自旋密度和Np-O_(yl)键长的结果也证实了乙醛肟还原Np(Ⅵ)的本质。本研究解释了Z/E-CH_(3)CHNOH还原Np(Ⅵ)的微小差异,并揭示了其还原本质,为乏燃料中镎的分离提供了理论依据和支持。

硫化铋修饰沸石对高温氩气环境中碘的净化研究

乏燃料后处理过程中产生的放射性气体中含有^(129)I,因其半衰期长、含量高且毒性大而备受关注。现阶段对干法后处理中^(129)I高温氩气环境条件下的研究很少,为探究固体多孔吸附剂在干法高温氩气环境中对放射性碘的吸附性能,本文以Bi(NO_(3))_(3)·5H_(2)O为铋源、L-半胱氨酸为硫源、乙二醇为溶剂,采用水热法制备硫化铋修饰沸石复合材料(Bi_(2)S_(3)-MOR),并采用该复合材料进行静态吸附实验。结果表明:在130℃氩气环境中,硫化铋修饰沸石对单质碘的静态吸附容量可达180 mg/g,捕集形式既有化学吸附的BiI_(3),又有物理吸附的I_(2),而在50℃氩气环境中,对甲基碘的静态吸附容量为50 mg/g,仅存在BiI_(3)形式的化学吸附。水热法Bi_(2)S_(3)修饰沸石表现出同Ag^(0)修饰沸石一样优异的碘吸附性能。

核电厂乏燃料水池扩容改造辐射防护技术实践与发展

在核电厂乏燃料水池扩容改造过程中,辐射防护工作至关重要。项目组严格按照安全标准制定辐射防护方案,从培训、模拟实操练习、剂量限值控制、辐射防护管理和防护措施、个人剂量监测、工作场所监测、工艺监测、突发事件的应急措施等方面作了全方位的准备,严格按照程序执行工作,为乏燃料水池扩容改造的顺利完成提供辐射防护保障。

Passive neutron multiplicity device for^(240)Pu measurement based on FPGA

A passive neutron multiplicity measurement device,FH-NCM/S1,based on field-programmable gate arrays(FPGAs),is developed specifically for measuring the mass of plutonium-240(^(240)Pu)in mixed oxide fuel.FH-NCM/S1 adopts an inte-grated approach,combining the shift register analysis mode with the pulse-position timestamp mode using an FPGA.The optimal effective length of the^(3)He neutron detector was determined to be 30 cm,and the thickness of the graphite reflector was ascertained to be 15 cm through MCNP simulations.After fabricating the device,calibration measurements were per-formed using a^(252)Cf neutron source;a detection efficiency of 43.07%and detector die-away time of 55.79μs were observed.Nine samples of plutonium oxide were measured under identical conditions using the FH-NCM/S1 in shift register analysis mode and a plutonium waste multiplicity counter.The obtained double rates underwent corrections for detection efficiency(ε)and double gate fraction(f_(d)),resulting in corrected double rates(D_(c)),which were used to validate the accuracy of the shift register analysis mode.Furthermore,the device exhibited fluctuations in the measurement results,and within a single 20 s measurement,these fluctuations remained below 10%.After 30 cycles,the relative error in the mass of^(240)Pu was less than 5%.Finally,correlation calculations confirmed the robust consistency of both measurement modes.This study holds specific significance for the subsequent design and development of neutron multiplicity devices.

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。

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。

选区激光熔化Ti-6Al-4V合金在模拟乏燃料后处理环境中的腐蚀行为

研究选区激光熔化(SLM)制备的Ti-6Al-4V合金在乏燃料后处理环境中的腐蚀行为。借助光学显微镜、X射线衍射、扫描电子显微镜、电子背散射衍射以及电化学测试等手段研究SLMTi-6Al-4V合金的显微组织及耐蚀性。结果表明,相较于铸造合金,SLM Ti-6Al-4V合金在含氧化性离子的6 mol/L热硝酸溶液中的耐腐蚀性更好。进一步分析表明,SLMTi-6Al-4V合金具有α'+β的微观结构,β-Ti相含量较高且分布均匀,同时其晶粒尺寸较小、晶界密度较高,这种微观结构可提高其在含氧化性离子的硝酸中的钝化能力以及耐腐蚀性。SLM Ti-6Al-4V合金有望在乏燃料后处理过程中得到广泛应用。