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 flexi-bility 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 nu-clear 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 poower, Because of the inherent flexibility of nuclearfuel 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 modifica-tion 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.     

PWR/CANDU联合核燃料循环研究

根据我国已拥有PWR和CANDU核电站的具体情况 ,提出一种PWR/CANDU联合核燃料循环的策略 ,即把压水堆的乏燃料后处理后的回收铀 (RU)用作为CANDU堆的核燃料 ,既可节约铀资源 ,提高燃料的能量输出 ,又减少了废燃料的处置量 ,可大大降低核电成本。由于CANDU堆对核燃料循环的固有灵活性 ,堆芯结构及运行方式不需作重大改变 ,即可完成从天然铀到RU的过渡。又由于RU较低的放射性活度 ,这对CANDU堆的燃料制造是可以接受的 ,因而只需对现有燃料制造生产线稍加屏蔽措施 ,对运输和运行中燃料管理操作等都勿须改变。因而这一策略是具有重大经济效益和吸引力的     

Comparative cost analysis of direct disposal versus pyro-processing with DUPIC in Korea

There are 20 nuclear reactors operating in Korea and four more are under construction. The spent nuclear fuel and radioactive wastes will be accumulated and an effective management must be introduced. In Korea, pyro-processing and Direct Use of spent PWR fuel In CANDU (DUPIC) are drawing attentions of many researchers and policy makers. However, cost comparison studies of each or both options against the direct disposal have not been adequately conducted. Therefore, the purpose of this study is to compare the fuel cycle strategy in cost terms. Based on mass balance of the Heavy Metal, cost of each process in the fuel cycle was considered. It was found that the pyro-processing-only cannot win against the direct disposal, but it can be compensated by adding DUPIC. Pyro-processing with DUPIC was cheaper by 925 M$ than the direct disposal. Further time-considered analysis will supplement this work with reasonable basis.     

Fire in the hole: A review of national spent nuclear fuel disposal policy

Today, nuclear power plants operate in 31 countries and account for approximately 15% of the world's electricity production. Moreover, a number of additional reactors are expected to come on-line over the next several years as a result of a global resurgence in nuclear power. Since the amount of spent nuclear fuel generated by these plants is expected almost to double by 2020, the issue of how to dispose of highly radioactive waste properly is an international concern of growing importance. As of now, no country has yet solved the problem of what to do with the mounting inventories of spent nuclear fuel created as a by-product of nuclear power generation. This article focuses on national approaches for the disposal of spent nuclear fuel, and discusses the need for a global approach to cope adequately with the increasing inventories of highly radioactive waste.     

Analysis of thermal problems in development of spent fuel transport cask for research reactors

A packaging for the transport of irradiated fuel from research reactors was designed by a group of researchers to improve the capability in the management of spent fuel elements from the reactors operated in the region. Two half scale models for MTR fuel were constructed and tested so far and a third one for both MTR and TRIGA fuels will be constructed and tested next. Four test campaigns have been carried out, covering both normal and hypothetical accident conditions of transportation. The thermal test is part of the requirements for the qualification of transportation packages for nuclear reactors spent fuel elements. In this paper, both the numerical modelling and experimental thermal tests performed are presented and discussed. The cask is briefly described as well as the finite element model developed and the main adopted hypotheses for the thermal phenomena. The results of both numerical runs and experimental tests are discussed as a tool to validate the thermal modelling. The impact limiters, attached to the cask for protection, were not modelled.     

Current and Future Problems of the Fuel Supply for Nuclear Power

The structure of the nuclear fuel cycle, consisting of the technological stages of uranium production, refining, enrichment, fabrication of nuclear fuel, and reprocessing of the spent fuel for reuse of the fissioning materials, is examined. Supplying fuel includes supplying fuel for Russian nuclear power plants, propulsion and research reactors, export of fuel for nuclear power plants and research reactors constructed according to Russian designs, export of low-enriched uranium and fuel for nuclear power plants constructed according to foreign designs. The explored deposits of natural uranium, the estimated stores of uranium in reserve deposits, and warehoused stores will provide nuclear power with uranium up to 2030 and in more distant future with the planned rates of development. The transition of nuclear power plants to a new fuel run will save up to 20% of the natural uranium. The volume of reprocessing of spent fuel and reuse of ²³⁵U makes it possible to satisfy up to 30% of the demand for resources required for Russian nuclear power plants. The most efficient measure of the resource safety of Russian nuclear power is implementation of an interconnected strategy at each stage of the nuclear fuel cycle.     

Possibilities of BREST reactors and transmutation fuel cycle under conditions implementing modern plans for the development of nuclear power

The possible dynamics of the development of BREST-1200 fast reactor capacities after 2030 on the basis of plutonium and other actinides accumulated in the spent fuel of thermal reactors is examined. It is shown that by 2100 the power BREST reactors could be 114–176 GW, and subsequently they will develop as a result of their own breeding of plutonium. Calculations have shown that the rate at which BREST reactors are put into operation can be doubled by using enriched uranium obtained from natural uranium and regenerated spent fuel from thermal reactors. It is shown that the development of fast reactors with a closed fuel cycle solves the problem of transmutation of long-lived high-level actinides and makes it possible to implement a transmutation fuel cycle in nuclear power. __________ Translated from Atomnaya énergiya,Vol. 103, No. 1, pp. 21–28, July, 2007.     

Aging management and post-storage transport of dual purpose cask for spent fuel

Aging management of spent fuel storage facility may follow lessons learned from literature for nuclear power plant and a review for spent fuel dry cask storage system by US NRC, DOE, by German BAM, that by Japan NISA, etc. Namely, the essence of systematic approach to aging management includes Understanding aging, Plan (Development and optimisation of activities for aging management), Do (Managing aging mechanisms), Check (Monitoring, inspection and assessment), and Act (Maintenance). The PDCA cycle will optimise the systematic approach to the aging management. An aging management programme (AMP) for the storage system over the period of extended storage will address uncertainties in the safety relevant functions of the system that may otherwise be impaired by aging mechanisms. The AMP identifies system, structure and components (SSCs) that need specific actions to mitigate aging and ensures that no aging effects result in a loss of their intended function of the SSCs, during an intended licensed period. AMPs generally include Prevention, Mitigation, Monitoring, Inspection, and Maintenance programmes. Aging management plans should ensure compliance with transportation requirements after extended storage. Potential issue would be a significant change of the transport regulations in the future. If the regulations changed significantly, a gap analysis should be performed to identify any impact to the cask safety. Compensating arrangements, if necessary, should be proposed at that time. Assuming that the regulations will not change significantly after long term storage, we will be able to renew the license both for transport and storage of the cask during the storage period. For example, in Japan, a holistic approach was established for the license of a 50 year storage and transport. In this approach, we can evaluate integrity of spent fuel, basket, etc. with respect to chemical, thermal, mechanical, and radiation factors. With this approach we will not have to open the cask lid for visual inspection of the spent fuel, basket, etc. prior to the post-storage transport.     

Transmutation of Americium in a Heavy-Water Reactor

It is shown that 22.5 metric tons of americium from the spent fuel of 30 VVÉR reactors which operated for 30 yr can be transmuted in a 1 GW(t) heavy-water system in 103 yr using as fuel the plutonium from the same spent VVÉR fuel. This means that 7.5 VVÉR reactors (CUF = 0.85) must be maintained simultaneously for fuel storage time 30 yr (for a 3-yr fuel storage period, the number of VVÉR reactors maintained increases to 25). In the entire period of operation of the system, a substantial quantity of plutonium from the spent fuel is used – about 150 metric tons (with total plutonium production in VVÉR reactors of about 200 metric tons) and about 38 metric tons of fissioning isotopes are burned. Therefore, with up to 98% burnup of americium in the target material the conversion coefficient defined as the ratio of the mass of the americium annihilated to the mass of the spent fissioning material is about 0.57.     

乏燃料水池严重事故现象及管理策略研究

针对我国二代改进型三环路核电厂乏燃料水池冷却管线破口事故(LOCA)引发的严重事故,使用MECLOR1.8.6程序进行了建模计算,分析研究了严重事故进程和乏燃料组件加热、熔化以及氢气的产生等主要现象。结果表明,乏燃料水池严重事故进程相对缓慢,但乏燃料组件的熔化及产生的氢气风险还是可能最终造成放射性向环境的大量释放。此外,本文还对乏燃料水池严重事故管理导则中的应急注水策略和氢气风险管理策略的有效性进行了计算分析,得到了严重事故下执行相关策略的时间窗口,从而为同类型核电厂严重事故管理导则的开发和有效执行提供支持。     

Systems Analysis of Spent Fuel Management in Japan, (I)

Spent fuel accumulation in the future and its appropriate management strategy for Japan is analyzed by “SFTRACE” (Spent Fuel Storage, TRAnsportation and Cost Evaluation System) is introduced, which consists of 3 sub-models; (1) the economic cost data base for spent fuel storage technologies, (2) the long-range simulation of reactor mix and plutonium (Pu) utilization, and (3) the detailed simulation of spent fuel management strategies.

The long-range simulation sub-model presents a macroscopic overview on how much amount of spent fuel stockpile should be addressed nationwide at a certain time point. A preliminary calculation shows that the spent fuel storage needs in Japan to the year 2050 will vary significantly, from a decrease towards zero or a continuous increase up to 20,000–25,000 tHM. The sub-model for spent fuel management simulation is the tool to demonstrate nationwide strategies to deal with spent fuel accumulation, either at each power station site or a number of centralized facilities in the given time horizon, with associated needs of transportation. An illustrative analysis shows trade-off relationship among factors involved in spent fuel management strategies, such as between “away from reactor (AFR)” storage capacity and overall transportation requirements, which vividly demonstrates the usefulness of the integrated analytic tool.     

“CANDLE” burnup regime after LWR regime

CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) burnup strategy can derive many merits. From safety point of view, the change of excess reactivity along burnup is theoretically zero, and the core characteristics, such as power feedback coefficients and power peaking factor, are not changed along burnup. Application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replacing fuels. About 40% of natural or depleted uranium undergoes fission without the conventional reprocessing and enrichment.

If the LWR produced energy of X Joules, the CANDLE reactor can produce about 50X Joules from the depleted uranium left at the enrichment facility for the LWR fuel. If we can say LWRs have produced energy sufficient for full 20 years, we can produce the energy for 1000 years by using the CANDLE reactors with depleted uranium. We need not mine any uranium ore, and do not need reprocessing facility. The burnup of spent fuel becomes 10 times. Therefore, the spent fuel amount per produced energy is also reduced to one-tenth.

The details of the scenario of CANDLE burnup regime after LWR regime will be presented at the symposium.     

Estimation of Spent Fuel Compositions from Light Water Reactors

The compositions and quantities of minor actinide (MA) and fission product (FP) in spent fuels will be diversified with the use of high discharged burnup fuels and MOX fuels in LWRs which will be a main part of power reactors in future.

In order to investigate above diversities, we have studied on the calculation method to be used in the estimation of spent fuel compositions and adopted the real irradiation calculation in which axial burnup and moderator distribution are considered in the burnup calculation.

On the basis of the calculations, compositions and burnup quantities of various LWR spent fuels (reactor type: PWR and BWR, discharged burnup: 33, 45 and 60 GWd/tHM, fuel type: U0₂ and MOX) are apparently estimated among various forms of fuels. As an example, it is shown that there are considerable discrepancy in MA burnup between PWR and BWR spent fuels.     

Innovative development of nuclear power in Russia

The basic principles for performing analysis and the systems requirements for large-scale nuclear power in our country are formulated. The problems of modern nuclear power are examined and ways for modern nuclear power to transition to innovative development while satisfying these systems requirements for fuel use, handling spent fuel and wastes, and nonproliferation are indicated. The basic scenario of innovative development in the near term (up to 2030) is based on using predominantly ²³⁵U as fuel and water-moderated water-cooled reactors, which have been well mastered, for increasing nuclear capacities with limited introduction of fast reactors for solving the problem of spent fuel from thermal reactors. In the long term (2030–2050), a transition to ²³⁸U as the primary raw material with fast reactors predominating and complete closure of the nuclear power fuel cycle will be made. The journal variant of a report “New-Generation Nuclear Energy Technologies” presented at a meeting of the Scientific and Technical Council of Rosatom, Moscow, September 27, 2006. __________ Translated from Atomnaya énergiya, Vol. 103, No. 3, pp. 147–155, September, 2007.     

2020年前我国核燃料循环情景初步研究

根据我国核电现状和中短期发展规划,对2020年前我国核电规模提出了三种预测方案,并根据各种方案对压水堆电站的核燃料循环情景进行了计算。重点研究了压水堆核电所需的铀资源、分离功,卸出的乏燃料以及乏燃料中Pu和次要锕系元素(MA)的产生量。     

Recycling of uranium from spent RBMK fuel in the nuclear fuel cycle

It is shown that there is promise in using the uranium product obtained by reprocessing spent nuclear fuel from RBMK reactors as a non-initial fuel source for thermal reactors. A technical path for spent nuclear fuel from RBMK reactors is proposed: radiochemical reprocessing and obtaining oxides of recycled uranium. Oxides of the category RBMK-poor are packed and then stored in a near-surface storage facility; oxides of the category RBMK-rich are fluoridated, and UF₆ is fed into separation production for additional enrichment to the required content of ²³⁵U. Additional advantages of recycled RBMK uranium as a source of non-initial ²³⁵U are the low content of ²³²U and the relatively low activity of spent fuel, which simplifies its reprocessing.     

VALMOX: validation of nuclear data for high burn-up MOX fuels

VALMOX, an acronym for validation of nuclear data for high burn-up MOX fuels, is one of the projects of the cluster evolutionary fuel concepts: high burn-up and MOX fuels (EVOL). It covers 30 months, from October 2001 to March 2004.It considers the evaluation of the actinide inventory of MOX fuel at high burn-up (typically 60 GWd/t) in light water reactors, with special attention to the helium production. Calculated values for the spent fuel isotopic masses are compared to the measured ones, with sensitivity analyses made in support. The JEF 2.2 nuclear data file is taken as a basis for calculation. The resulting recommendations on nuclear data should be employed in the preparation and testing of the next JEFF3 file.So far, the major effort was placed on the evaluation of MOX fuel irradiations in pressurised water reactors, and first results will be presented and compared.     

Plutonium management with thorium-based fuels and inert matrix fuels in thermal reactor systems

The plutonium that is produced by light water reactors worldwide is currently re-used to a limited extent. In the last century, the expected introduction of fast reactors and the associated need for large amounts of plutonium did not take place. The result is that worldwide a stockpile of excess plutonium has formed, which is the dominant contributor to the radiotoxicity of spent nuclear fuel for storage times from 10² to 10⁵ years. One option to reduce or stabilize the plutonium stockpile is to utilize this plutonium in advanced fuel types, such as thorium-based and inert matrix fuels. Because these fuels do not contain uranium, the plutonium consumption rate is very high. In this paper, the status of the fuel research and some recent developments are given.     

The reactor physics characteristics of a transuranic mixed oxide fuel in a heavy water moderated reactor

The reprocessing actinide materials extracted from spent fuel for use in mixed oxide fuels is a key component in maximizing the spent fuel repository utility. While fast spectrum reactor technologies are being considered in order to close the fuel cycle, and transmute these actinides, there is potential to utilize existing pressurized heavy water reactors such as the CANDU^®1 design to meet these goals. The use of current thermal reactors as an intermediary step which can burn actinide based fuels can significantly reduce the fast reactor infrastructure needed. This paper examines the features of actinide mixed oxide fuel, TRUMOX, in a typical CANDU nuclear reactor. The actinide concentrations used were based on extraction from 30 year cooled spent fuel and mixed with natural uranium in 4.75% actinide MOX fuel. The WIMS-AECL model of the fuel lattice was created and the two neutron group properties were transferred to RFSP in order to create a 3 dimensional time average full core model. The model was created with typical CANDU limits on bundle and channel powers and a burnup target of 45 MWd/kgHE. The TRUMOX fuel design achieved its goals and performed well under normal operations simulations. This effort demonstrated the feasibility of using the current fleet of CANDU reactors as an intermediary step in burning reprocessed spent fuel and reducing actinide burdens within the end repository. The recycling, reprocessing and reuse of spent fuels produces a much more sustainable and efficient nuclear fuel cycle using existing and proven reactor technologies.     

The history and the prospects of japanese nuclear power development program

FBRs are regarded as the most probable option among non-fossil energy resources which will underpin the future energy demand in Japan, considering the effective uranium utilization and the need to lower the burden on the natural environment. However, it will take a long time to utilize FBRs due to a number of pending technical issues and improvements of cost efficiency. For the time being, therefore, light water reactors will continue to play a dominant role in power generation: thus, it is urgently necessary to establish the quasi-domestic nuclear fuel cycle for them, especially in the field of enrichment and spent fuel reprocessing — a goal of the Japanese nuclear policy since the dawn. Furthermore, public acceptance is significant factor which must be considered. This can best be achieved by more safety performance of light water reactors and through publication of extensive information, including decisions by the industry and government.