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.     

Plutonium and Minor Actinides incineration options using innovative Na-cooled fast reactors: Impacting on phasing-out and on-going fuel cycles

The flexibility of innovative Na-cooled fast reactors for burning Pu and/or Minor Actinides (MA) is investigated with respect to different fuel cycle strategies. Under phasing-out conditions, the burner systems are used for reducing to a minimum level the accumulated TRansUranic (TRU) inventory, whereas when continuous use of nuclear energy is envisaged (on-going case), burner systems may be dedicated to MA management only.As an example of a phasing-out case, the accumulated German TRU inventory (at 2022) is assumed to be transmuted in a chosen time period of 150 years. For this purpose, two different burner fast reactors concepts, developed at KIT, are deployed in a Partitioning and Transmutation based fuel cycle. The effects are analyzed in order to confirm the behavior expected by the neutronics studies and to provide a basis for further optimization of the scenarios with respect to a number of reactors, deployment paces and fuel compositions.Additionally the performance of the MA burner is assessed to provide an effective MA mass stabilization in case of a continuous use of nuclear energy. Preliminary results are compared with those of past studies based on the European Sodium-cooled Fast Reactor.     

Practice and prospect of advanced fuel management and fuel technology application in PWR in China

Since Daya Bay nuclear power plant implemented 18-month refueling strategy in 2001, China has completed a series of innovative fuel management and fuel technology projects, including the Ling Ao Advanced Fuel Management (AFM) project (high-burnup quarter core refueling) and the Ningde 18-month refueling project with gadolinium-bearing fuel in initial core. First, this paper gives brief introduction to China's advanced fuel management and fuel technology experience. Second, it introduces practices of the advanced fuel management in China in detail, which mainly focuses on the implementation and progress of the Ningde 18-month refueling project with gadolinium-bearing fuel in initial core. Finally, the paper introduces the practices of advanced fuel technology in China and gives the outlook of the future advanced fuel management and fuel technology in this field.     

Fabrication of DUPIC fuel pellets using high burn-up spent PWR fuel

Technology for the direct usage of a spent PWR fuel in CANDU reactors (DUPIC) was developed in KAERI to reduce the amount of spent fuel. DUPIC fuel pellets were fabricated using a dry processing method to re-fabricate CANDU fuel from spent PWR fuel without any intentional separation of fissile materials or fission products. The DUPIC fuel element fabrication process satisfied a quality assurance program in accordance with the Canadian standard. For the DUPIC fuels with various fuel burn-ups between 27,300 and 65,000 MWd/tU, the sintered pellet density decreased with increasing fuel burn-ups. Fission gas releases and powder properties of the spent fuel also influenced the DUPIC fuel characteristics. Measurement of cesium content released from green pellets revealed that their sintered density significantly depended on sintering temperature history. It was useful to establish a DUPIC fuel fabrication technology in which a high-burn-up fuel with 65,000 MWd/tU was treated.     

Performance of a transmutation advanced device for sustainable energy application

Preliminary studies have been performed to design a device for nuclear waste transmutation and hydrogen generation based on a gas-cooled pebble bed accelerator driven system, TADSEA (Transmutation Advanced Device for Sustainable Energy Application). In previous studies we have addressed the viability of an ADS Transmutation device that uses as fuel wastes from the existing LWR power plants, encapsulated in graphite in the form of pebble beds, cooled by helium which enables high temperatures (in the order of 1200 K), to generate hydrogen from water either by high temperature electrolysis or by thermochemical cycles. For designing this device several configurations were studied, including several reflectors thickness, to achieve the desired parameters, the transmutation of nuclear waste and the production of 100 MW of thermal power. In this paper new studies performed on deep burn in-core fuel management strategy for LWR waste are presented. The fuel cycle on TADSEA device has been analyzed based on both: driven and transmutation fuel that had been proposed by the General Atomic design of a gas turbine-modular helium reactor. The transmutation results of the three fuel management strategies, using driven, transmutation and standard LWR spent fuel were compared, and several parameters describing the neutron performance of TADSEA nuclear core as the fuel and moderator temperature reactivity coefficients and transmutation chain, are also presented.     

Project SPHINX spent hot fuel incinerator by neutron flux (the development of a new reactor concept with liquid fuel based on molten fluorides)

The SPHINX project is dealing with a solution of some principle problems of a very promising way of nuclear waste treatment, high level wastes from spent nuclear fuel in particular, by means of transmutation of radionuclides by use of a nuclear reactor with liquid fuel based on molten fluorides, which might be a subcritical system driven by a suitable neutron source. Its superiority lies also in the fact that it makes possible to utilize actinides contained, by others, in spent nuclear fuel and so to reach a positive energy effect.

The SPHINX project has been proposed by the consortium TRANSMUTATION being established by four leading nuclear research bodies in the Czech Republic (Nuclear Research Institute Rez plc, SKODA Nuclear Machinery plc in Pilsen, Nuclear Physics Institute of Academy of Sciences in Rez and Technical University in Praha) at the end of 1996 to which Technical University in Brno (specialized for a secondary circuit problems) has associated in the year 2000. The project has been supported by the Ministry of Industry and Trade of the Czech Republic, CEZ, a.s. (Czech Electricity Generating Company) and RAWRA (Radwaste Repository Authority).

The R&D program of the SPHINX project contains an experimental part, which serves for a verification of design inputs for designing a demonstration unit of a transmuter with liquid fuel based on molten fluorides. The current status of the experimental program performance has been focused upon the irradiation of samples of molten-salt systems as well as structural materials proposed for the blanket of the SPHINX transmuter in the field of high neutron flux of research reactors.

The main aims of this program called Irradiated Probes BLANKA are the following: (1) Experimental verification of long time behavior of transmuter blanket which contains molten fluoride salts as a fuel and coolant, (2) Validation of computational code system being developed for the computation of actinides concentration in long- term operation of the transmuter, and (3) Material research on behavior of materials in neutron and gamma fields, and materials interactions on high temperature conditions.

At present, two agreements on multinational cooperation in this field have been signed: One with European Commission and one with Russian Kurchatov Institute (joint experimental programs AMPULA containing fluorides of transuranium elements like Np, Pu, Am and Cm in irradiated samples and a joint development of the ISTAR code).     

Current status of spent fuels and the development of computer programs for the PWR spent fuel management in Korea

Nuclear power has supplied the national electric power demand for three decades in the Republic of Korea, which has resulted in the accumulation of a large amount of spent fuels. The government has a policy on the temporary storage of these at nuclear power plants at present. In order to establish a proper policy for spent fuel management in the near future, the characteristics and amount of spent fuels should be figured out properly. In this paper, the current status of spent fuels in the Republic of Korea is outlined focusing on the major characteristics of spent fuels such as initial enrichment and discharge burnup. According to the current trend, the average burnup of PWR spent fuels will reach 55 GWd/MtU by the middle of 2010s. Three different kinds of computer programs were developed to supply crucial data regarding spent fuels. The first one was developed to project the amount of spent fuels in the future based on three different projection models. The projection was verified with real spent fuel data. The second Database program was prepared for the analysis of statistics regarding PWR spent fuels. Each PWR spent fuel assembly was specified with 18 items of data such as fuel type, initial enrichment, and discharge burnup. The usefulness of the Database program was illustrated through an analysis of the geological disposal density and cooling time of PWR spent fuels. Disposal area could be reduced by 50% through a proper analysis of the cooling time of PWR spent fuels. Finally, A-SOURCE program was developed to easily calculate source-terms such as decay heat and radionuclide concentration after the pyro-processing of PWR spent fuel assemblies. Linked to the Database program, the A-SOURCE program selected PWR spent fuel assemblies and could calculate the source-terms for any combination of them. An illustration of the usage of the program was demonstrated.     

Isotope correlation technique for Pu mass analysis of PWR UO2 spent fuel rods

A simple and fast method of nuclear material accountancy of pressurized water reactor (PWR) UO₂ spent fuel rods for safeguards application was developed utilizing the isotope correlation between the amounts of ¹³⁷Cs and total Pu. To this end, the following steps were taken: (1) as much destructive analysis (DA) data as possible for segments taken from a PWR UO₂ spent fuel rod were aggregated from publicly available data sources; (2) the DA data were corrected so as to have the same cooling time (i.e., CT = 0 y) and analyzed for outliers; (3) an equation converting the ¹³⁷Cs amount to the Pu amount was obtained by regression analysis with logarithmic curve fitting; and (4) the error in determining the Pu amount was evaluated for the imposition of a limit on the range of burnup (BU) or initial enrichment (IE). It was found that the averaged % error in calibration was determined to be 3.88% ± 2.68% (= mean ± 1 standard deviation) for the BU range over 30 GWd/tU and falling with increasing BU range. On the other hand, there was no benefit in applying the limit of the IE range. Lastly, the Pu-mass difference between various methods was compared and it was found that the difference can be incurred up to 11.4%, according to the choice of method. In conclusion, the proposed isotope correlation technique could be used for input material accountancy with reasonable uncertainty.     

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。     

Measurement of homogeneity of U and Pu isotopes in PWR spent fuel powder mixture

We present a feasibility study of the homogenization of pressurized water reactor spent nuclear fuel (SNF) powder through a mechanical mixing process. Because burn-up of the SNF depends on the position in the SNF assembly, concentrations of uranium, plutonium, and fission products are distributed differently according to the burn-up profile. The heterogeneity of the material elements affects the selection of a representative sample for quantitative analysis. Homogenization process improvement to reduce the sampling error is thus required to precisely determine the amount of uranium, plutonium, and fission products in the SNF. In this study, fine powders (<70 μm) extracted from one SNF rod were mixed, and the degree of homogenization was determined as a function of the mixing time indicating the relative standard deviations of the ¹³⁴Cs/¹³⁷Cs, Pu, and U isotope ratio measurements.     

Development of an advanced PWR fuel for OPR1000s in Korea

The advanced PWR fuel for the OPR1000s in Korea, PLUS7, has been developed to enhance thermal performance, high burnup capability and fuel reliability against grid-to-rod fretting wear and debris. The outstanding design features of PLUS7 include mixing vane mid-grids for increasing thermal performance and minimizing vibration-induced fretting wear, optimized fuel dimensions and advanced zirconium alloys for high burnup capability of 72,000 MWD/MTU, and an optimized fuel rod diameter for reducing pressure drop and improving neutron economy. The fuel assembly and its components performances have been verified through a wide spectrum of mechanical, thermal hydraulic, vibration and fretting wear tests. Based on the verification test results and the evaluations with the help of the KNF design code system, it is found that the PLUS7 fuel will maintain its integrity up to the envisaged burnup of 72,000 MWD/MTU. In addition, the PLUS7 fuel performances were evaluated to be considerably improved in comparison with the current fuel used in the OPR1000s.     

Evaluation of uranium thorium and plutonium thorium fuel cycles in a very high temperature hybrid system

In recent times, there is a renewed and additional interest in thorium because of its interesting benefits. Thorium fuel cycle is an attractive way to produce long term nuclear energy with low radiotoxicity waste. In addition, the transition to thorium could be done through the incineration of weapons grade plutonium or civilian plutonium. Th-based fuel cycles have intrinsic proliferation-resistance and thorium is 3–4 times more abundant than uranium. Therefore, thorium fuels can complement uranium fuels and ensure long term sustainability of nuclear power.In this paper, the main advantages of the use of fuel cycles based on uranium-thorium and plutonium-thorium fuel mixtures are evaluated in a hybrid system to reach the deep burn of the fuel. To reach this goal, the preliminary conceptual design of a hybrid system composed of a critical reactor and two Accelerated Driven Systems, of the type of very high temperature pebble-bed systems, moderated by graphite and cooled by gas, is analyzed.Uranium-thorium and plutonium-thorium once-through and two stages fuel cycles are evaluated. Several parameters describing fuel behaviour and minor actinide stockpile are compared for the analyzed cycles.     

Progress in the studies of the advanced plutonium fuel assembly

A new assembly concept, designated APA (for dvanced lutonium fuel ssembly), should make it possible to multi-recycle plutonium in pressurized water reactors. The basic idea is founded on the manufacture of a large plutonium thin annular fuel rod with an inert support, cooled on both faces. The absence of plutonium generation, combined with moderate fuel temperature should make it possible to achieve substantial burn-up fractions in these rods. The assembly is compatible with the internals of a Pressurized Water Reactor (PWR), and provides for permanent reversibility. Neutronic studies showed a compliance with actual safety/control criteria. A multi-recycling scenario was simulated for 84 years' operation with a 65 GW electrical power installed capacity, comprising forty-five 1450 MW electrical power PWRs, 32 of which are loaded with UO₂ and 13 with APAs. It showed that the plutonium inventory is controlled. Thermal-hydraulic studies showed one can find an annular rod geometry allowing one to respect both margins to Critical Heat Flux (CHF) during normal and accidental operations and void fraction limitations.     

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.     

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.     

Consideration of aging mechanism influence on transport safety of dual purpose casks for spent nuclear fuel or HLW

When storage of spent nuclear fuel or high level waste is carried out in dual purpose casks (DPC), the effects of aging on safety relevant DPC functions and properties have to be managed in a way that a safe transport after the storage period of several decades is capable and can be justified and certified permanently throughout that period. The effects of aging mechanisms (e.g. radiation, different corrosion mechanisms, stress relaxation, creep, structural changes and degradation) on the transport package design safety assessment features have to be evaluated. Consideration of these issues in the DPC transport safety case will be addressed. Special attention is given to all cask components that cannot be directly inspected or changed without opening the cask cavity, like the inner parts of the closure system and the cask internals, like baskets or spent fuel assemblies. The design criteria of that transport safety case have to consider the operational impacts during storage. Aging is not the subject of technical aspects only but also of ‘intellectual’ aspects, like changing standards, scientific/technical knowledge development and personal as well as institutional alterations. Those aspects are to be considered in the management system of license holders and in appropriate design approval update processes. The paper addresses issues that are subject of an actual International Atomic Energy Agency TECDOC draft ‘Preparation of a safety case for a dual purpose cask containing spent nuclear fuel’.     

乏燃料公海铁联运核应急体系建设研究

公海铁联运作为解决大宗乏燃料远距离运输的最佳方案,在国际上是一种较为普遍的运输模式,如果未来我国采用该运输模式,需探索相关核应急工作思路。本文调研梳理了国内外乏燃料公海铁联运核应急相关法规标准,参考借鉴国外乏燃料运输相关实践,提出我国乏燃料公海铁联运核应急体系建设相关工作建议。     

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。     

Dynamic assessments on high-level waste and low- and intermediate-level waste generation from open and closed nuclear fuel cycles in Republic of Korea

We estimated the generation of low- and intermediate-level waste (LILW) and high-level waste (HLW) from open and closed nuclear fuel cycles. The closed fuel cycle reflects the development and deployment of fast reactors and pyroprocessing from 2013 to 2100, while the open fuel cycle only considers pressurized water reactors. The closed fuel cycle hardly affects short-term spent fuel management but can save nearly 60% space of interim storage compared with the open fuel cycle. Compared with the open fuel cycle, the accumulated volume of HLW can be significantly reduced in the closed fuel cycle up to over 95% in 2100. For this volume reduction, high heat generating fission products should be separated from transuranic waste in pyroprocessing and stored in decay storages for a few hundred years. Mining and milling waste in the closed fuel cycle decreases by about 31%. In contrast, the closed fuel cycle generates 3.0%–4.5% more LILW than the open fuel cycle because fast reactors and pyroprocessing produce more LILW and conversion, enrichment, and fabrication produce less LILW. In the closed fuel cycle, operation and decommissioning wastes from reactor and pyroprocessing, respectively, contribute to 74% and 8% of LILW excluding mining and milling waste.     

Structure of ion tracks in ceria irradiated with high energy xenon ions

Structure and accumulation behavior of ion tracks in CeO₂ irradiated with 200 MeV Xe ions were examined by transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM) to obtain fundamental knowledge on the microstructure evolution induced by fission fragments in nuclear fuels and transmutation targets, which is of importance for the development of advanced fuel/target materials at high burn-up conditions. Bright-field (BF) TEM images of ion tracks from an inclined direction showed Fresnel contrast along penetrating path of incident ions. The signal intensity of high-angle annular dark-field (HAADF) STEM images was decreased at the core damage region of ion tracks along the path of ions, revealing the reduction of atomic density inside the ion track. Preferential formation of smaller and larger ion tracks was observed at a high ion fluence of 1 × 10¹⁴ cm⁻² compared to a low ion fluence of 1 × 10¹¹ cm⁻². Results were discussed due to the coalescences and incomplete recovery of the core damage regions during the overlap of high density electronic excitation damage, which is induced during the repetition of the formation and recovery of ion tracks within an influence region.