Radiolytic modelling of spent fuel oxidative dissolution mechanism. Calibration against UO2 dynamic leaching experiments

Calibration and testing are inherent aspects of any modelling exercise and consequently they are key issues in developing a model for the oxidative dissolution of spent fuel. In the present work we present the outcome of the calibration process for the kinetic constants of a UO₂ oxidative dissolution mechanism developed for using in a radiolytic model. Experimental data obtained in dynamic leaching experiments of unirradiated UO₂ has been used for this purpose. The iterative calibration process has provided some insight into the detailed mechanism taking place in the alteration of UO₂, particularly the role of OH radicals and their interaction with the carbonate system. The results show that, although more simulations are needed for testing in different experimental systems, the calibrated oxidative dissolution mechanism could be included in radiolytic models to gain confidence in the prediction of the long-term alteration rate of the spent fuel under repository conditions.     

Spent fuel radionuclide source-term model for assessing spent fuel performance in geological disposal. Part I: Assessment of the instant release fraction

A source-term model for the short-term release of radionuclides from spent nuclear fuel (SNF) has been developed. It provides quantitative estimates of the fraction of various radionuclides that are expected to be released rapidly (the instant release fraction, or IRF) when water contacts the UO₂ or MOX fuel after container breaching in a geological repository. The estimates are based on correlation of leaching data for radionuclides with fuel burnup and fission gas release. Extrapolation of the data to higher fuel burnup values is based on examination of data on fuel restructuring, such as rim development, and on fission gas release data, which permits bounding IRF values to be estimated assuming that radionuclide releases will be less than fission gas release. The consideration of long-term solid-state changes influencing the IRF prior to canister breaching is addressed by evaluating alpha self-irradiation enhanced diffusion, which may gradually increase the accumulation of fission products at grain boundaries.     

Radionuclide release from high burnup spent fuel during corrosion in salt brine in the presence of hydrogen overpressure

In the case of a contact between groundwater and Fe-based spent fuel disposal containers in a repository large amounts of hydrogen will be produced by the corrosion of iron, which may result in significant hydrogen pressures. To quantify to what extent the hydrogen overpressure may counteract radiolysis enhanced matrix dissolution, related experimental work has been performed. High burnup spent fuel was corroded in 5.6 mol (kg H₂O)⁻¹ NaCl solution applying H₂ overpressures (experimental set 1) <0.17 bar by radiolysis, (experimental set 2) 2.8 bar by Fe corrosion, (experimental set 3) 3.2 bar by external H₂ gas. In the absence of Fe (experimental set 3) the UO₂ matrix dissolution rate decreased by a factor of about 10. In this test the concentrations of U, Np, Tc in solution were found to be decreasing by at least two orders of magnitude, and ranging within the same level as in the presence of Fe powder (experimental set 2). However, Pu and Am concentrations (experimental set 3) were less affected, due to the high sorption capacity for these radioelements onto Fe corrosion products.     

Behaviour of spent HTR fuel elements in aquatic phases of repository host rock formations

One back-end option for spent HTR fuel elements proposed for future HTR fuel cycles in the EC is an open fuel cycle with direct disposal of conditioned or non-conditioned fuel elements. This option has already been chosen in Germany due to the political decision to terminate the use of HTR technology. First integral leaching investigations at Research Centre Juelich on the behaviour of spent HTR fuel in salt brines, typical of accident scenarios in a repository in salt, proved that the main part of the radionuclide inventory cannot be mobilised as long as the coated particles do not fail. However, such experiments will not lead to a useful model for performance assessment calculations, because a failure of the coatings by corrosion will not occur during experimental times of a few years. In order to get a robust and realistic model for the long-term behaviour in aqueous phases of host rock systems, it is necessary to understand the barrier function of the different parts of an HTR fuel element, i.e. the matrix graphite, the different coating materials, and the fuel kernel.Therefore, our attention is focused on understanding and modelling the barrier performance of the different parts of an HTR fuel element with respect to their barrier function, and on the development of an overall model for performance assessment. In order to understand this behaviour, it is necessary to start with investigations of unirradiated material, and to proceed with experiments with external gamma irradiation to determine the effects of oxidising radiolysis species. Further experiments with irradiated material have to be performed to investigate the influence of the irradiation damage, and finally an investigation has to be made of the irradiated material plus additional gamma irradiation. Experimental data are now available for the diffusive transport of radionuclides in the water-saturated graphite pore system, the corrosion rates of unirradiated graphite with and without external gamma irradiation and unirradiated and irradiated silicon carbide, and for the dissolution rates of UO₂ and (Th,U)O₂ fuel kernels with and without external gamma irradiation. All investigations were performed in aquatic phases from salt, granite, and clay host rock.     

Effect of helium on the microstructure of spent fuel in a repository: An operational approach

In a repository, the release of radionuclides from spent fuel rods will strongly depend on the pellet microstructure existing when water comes into contact with the spent fuel surface, i.e. after 10,000 years of disposal. During this period, a large quantity of He atoms is produced by α-disintegrations of actinides in the spent fuel. A conservative model is proposed here to evaluate the consequences of He on the spent fuel microstructure. According to the solubility and diffusion properties of He under repository conditions, two scenarios are considered: He atoms can be trapped in fission gas bubbles or form new bubbles. In spite of the conservative assumptions of the model, the calculated values of bubble or pore pressure are much lower than critical values derived from rupture criteria. No evolution of the microstructure of the spent UO₂ fuel is thus expected before the breaching of the canister.     

Modeling spent TRISO fuel for geological disposal: corrosion and failure under oxidizing conditions in the presence of water

High temperature gas reactors (HTGRs) are being considered for near term deployment in the United States under the GNEP program and farther term deployment under the Gen IV reactor design (U.S. DOE Nuclear Energy Research Advisory Committee, 2002). A common factor among current HTGR (prismatic or pebble) designs is the use of TRISO coated particle fuel. TRISO refers to the three types of coating layers (pyrolytic carbon, porous carbon, and silicon carbide) around the fuel kernel, which is both protected and contained by the layers. While there have been a number of reactors operated with coated particle fuel, and extensive amount of research has gone into designing new HTGRs, little work has been done on modeling and analysing the degradation rates of spent TRISO fuel for permanent geological disposal. An integral part of developing a spent fuel degradation modeling was to analyze the waste form without taking any consideration for engineering barriers. A basic model was developed to simulate the time to failure of spent TRISO fuel in a repository environment. Preliminary verification of the model was performed with comparison to output from a proprietary model called GARGOYLE that was also used to model degradation rates of TRISO fuel. A sensitivity study was performed to determine which fuel and repository parameters had the most significant effect on the predicted time to fuel particle failure. Results of the analysis indicate corrosion rates and thicknesses of the outer pyrolytic carbon and silicon carbide layers, along with the time dependent temperature of the spent fuel in the repository environment, have a significant effect on the time to particle failure. The thicknesses of the kernel, buffer, and IPyC layers along with the strength of the SiC layer and the pressure in the TRISO particle did not significantly alter the results from the model. It can be concluded that a better understanding of the corrosion rates of the OPyC and SiC layers, along with increasing the quality control of the OPyC and SiC layer thicknesses, can significantly reduce uncertainty in estimates of the time to failure of spent TRISO fuel in a repository environment.     

乏燃料处置的必要性及其处置库环境化学行为

为了确保核燃料循环的安全性,不宜处理的乏燃料也应该同玻璃固化体一样作为高放废物进行深地质处置。本文综述了一些前期工作,归纳了空气侵入和水的辐解产生氧化性产物是导致乏燃料UO₂基体氧化溶解的主要因素;核燃料浸出实验结果显示铀和锕系镧系元素每天的浸出量是相应核素总量的1/10⁷,比裂变产物的浸出速率小一个数量级。铁金属被各国选为高放废物处置容器材料的原因是其低价格、高强度和优秀的还原能力。在最不利的地下水侵入深地质处置库、近场处置容器防腐层破损的情景下,铁容器材料表面与地下水反应产生氢气,氢气通过还原反应消耗辐解产生的氧化性自由基和分子,并能还原乏燃料表面的U(Ⅳ),大幅度减缓乏燃料的腐蚀和溶解;乏燃料中裂变产物贵金属合金颗粒对氢气有催化作用;处置容器表面铁金属能还原沉积溶解的多价态核素U(Ⅵ)、Np(Ⅴ)、Tc(Ⅶ)、Se(Ⅳ)和Se(Ⅵ)。希望本文对我国确立以铁基金属为处置容器材料的包括乏燃料在内的高放废物深地质处置概念有参考作用。     

压水堆内钍-铀增殖循环研究——乏燃料特性分析

利用ORIGENS程序对压水堆钍基乏燃料的特性进行分析,揭示了钍基乏燃料在放射性毒性、衰变热、γ射线等方面的特性,相关结果可为钍基乏燃料的贮存、后处理和地质处置提供必要的参考。研究的乏燃料是压水堆内钍-铀增殖循环堆芯设计方案中的4种,包括UOX（铀氧化物）、MOX（钚铀混合氧化物）、PuThOX（钚钍混合氧化物）和U3ThOX（工业级²³³U-钍混合氧化物）。研究结果表明：1）由于超铀核素的含量极低,在卸料后1 000年内,U3ThOX的放射性毒性显著低于超铀核素含量高的乏燃料;2）由于²³²U衰变链中²⁰⁸Tl的贡献,钍基乏燃料中2.6 MeV能量附近的γ射线强度明显高于铀基乏燃料,而这一能量附近的γ射线强度在卸料后约10年达到局部峰值,所以,钍基乏燃料的后处理最好避开此时间。     

Simulations of H2O2 concentration profiles in the water surrounding spent nuclear fuel

A simple mathematical model describing the hydrogen peroxide concentration profile in water surrounding a spent nuclear fuel pellet as a function of time has been developed. The water volume is divided into smaller elements, and the processes that affect hydrogen peroxide concentration are applied to each volume element. The model includes production of H₂O₂ from α-radiolysis, surface reaction between H₂O₂ and UO₂ and diffusion. Simulations show that the surface concentration of H₂O₂ increases fairly rapidly and approaches the steady-state concentration. The time to reach steady-state is sufficiently short to be neglected compared to the times of interest when simulating spent fuel dissolution under deep repository conditions. Consequently, the steady-state approach can be used to estimate the rate for radiation-induced spent nuclear fuel dissolution.     

压水堆内钍-铀增殖循环研究——堆芯设计

在全UOX（铀氧化物）堆芯平衡循环的基础上,研究了UOX/PuThOX（钚钍混合氧化物）混合堆芯和UOX/U3ThOX（工业级²³³U-钍混合氧化物）混合堆芯的燃料管理方案设计,实现了钍 铀增殖循环。U3ThOX燃料组件在堆内停留6个燃料循环,平均循环长度较参考的全UOX堆芯增加5 EFPD;U3ThOX燃料组件卸料后冷却1年时易裂变核素存量较装料时增加了7%。为比较分析,设计了UOX/MOX（钚铀混合氧化物）混合堆芯的燃料管理方案。核特性分析结果表明：1）装载PuThOX燃料对堆芯核特性产生的影响与装载MOX燃料类似,硼微分价值和控制棒价值减小、临界硼浓度变大、慢化剂温度系数更负、停堆裕量减小、多普勒亏损更大;2） UOX/U3ThOX混合堆芯和参考的全UOX堆芯具备相似的核特性。     

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.     

Challenges in the renewal of the Swedish sea transport system

Abstract

Since 1985, SKB has successfully operated a sea transport system for transport of spent nuclear fuel and radioactive waste to the intermediate storage facility, Clab and the final repository, SFR, in Sweden. The main components in the system are the ship M/S Sigyn, transport casks for spent fuel and core components, IP2 containers and terminal vehicles.     

Impact of partitioning and transmutation on the high level waste management

Nuclear energy generates 30% of the electricity in the EU. Still, different countries of EU27 have very different attitudes towards the future use of nuclear energy for electricity generation or other uses. However, independently of the political decision of continuation or phase out of the nuclear energy, all countries using nuclear energy to generate electricity are facing the question of the final management of its spent nuclear fuel and other high level radioactive wastes.Partition and Transmutation are emerging technologies that when integrated in advanced fuel cycles can strongly influence on the final wastes from the nuclear industry and its management. A review of the progress on the understanding of their real potentialities and main conclusions from a large number of international studies are presented in this paper. In particular, the conclusions from the main NEA working groups and the EURATOM RED-IMPACT project are jointly discussed.In this paper the emphasis is put on the effects of Partitioning and Transmutation on the inventory reduction, the heat source reduction and its implications to the repository capacity and on the performance assessment of the final repository.     

压水堆核电厂乏燃料后处理平准成本分析

基于压水堆核电厂乏燃料后处理工厂的规模与建设费用的半定量关系,以年处理能力为800 t的商用后处理厂为例,采用自上而下的方法,重点分析了隔夜成本、建造期和建造期间的利率,对采用PUREX流程的后处理工厂的建设费用影响;并用与建造期相同的利率,计算了UOX乏燃料后处理的平准成本.     

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.     

Dynamic analysis of a thorium fuel cycle in CANDU reactors

The thorium fuel recycle scenarios through a Canada deuterium uranium (CANDU) reactor have been analyzed for two types of thorium fuel: homogeneous ThO₂UO₂ and heterogeneous ThO₂UO₂–DUPIC fuels. The recycling was performed with dry process fuel technology which has a proliferation resistance. For the once-through fuel cycle model, the existing nuclear power plant construction plan was considered up to 2016, while the nuclear demand growth rate from the year 2016 was assumed to be 0%. After setting up the once-through fuel cycle model, a thorium fuelled CANDU reactor was modeled to investigate the fuel cycle parameters. In this analysis, the spent fuel inventory as well as the amount of plutonium, minor actinides and fission products for the multiple recycling fuel cycle were estimated and compared to those of a once-through fuel cycle.     

Study of thorium advanced fuel cycle utilization in light water reactor VVER-440

The first step in investigation of thorium fuel is evaluation of the results obtained from the spectral code for this type of fuel. The benchmark summarized by IAEA in 2003 was used for partial validation of the code HELIOS 1.9. The benchmark was focused on a comparison of the methods and basic nuclear data. Acceptable results of benchmark comparison allowed examining and comparing different advanced nuclear fuel cycles under light water reactor conditions, especially in VVER-440. Cycles, calculations and results for VVER-440 reactors are presented in the paper. Two of the investigated thorium based fuels include one solely plutonium–thorium based fuel, while the other one is a plutonium–thorium based fuel with a content of reprocessed uranium. The third examined fuel cycle is a cycle with an inert-matrix fuel consisting of reprocessed plutonium and minor actinides (MA) fixed in an yttria-stabilized zirconium matrix. All of them are used to carry and burn or transmute plutonium created in the classical UOX cycle. The Pu transmutation rate and cumulating of Pu with MA in the spent fuel were compared mutually and with an UOX open cycle. The fuel cycle with an inert-matrix fuel was proven to be the best cycle for minimizing the production of Pu in the VVER-440 reactors.     

Incorporation of cerium and neodymium in uranyl phases

The potential for incorporating rare earth elements (REE) into/onto crystalline compounds has been evaluated by precipitating uranyl phases from aqueous solutions containing either cerium or neodymium. These REEs serve both as monitors for evaluating the potential repository behavior of REE radionuclides, and as surrogate elements for actinides (e.g., Ce⁴⁺ and Nd³⁺ for Pu⁴⁺ and Am³⁺, respectively). The present experiments examined the behavior of REE in the presence of ianthinite , becquerelite (Ca(UO₂)₆O₄(OH)₆(H₂O)₈), and other uranyl hydroxide compounds commonly noted as alteration products during the corrosion of UO₂, spent nuclear fuel, and naturally occurring uraninite. The results of these experiments demonstrate that significant quantities of both cerium (K_d = 1020) and neodymium (K_d = 840) are incorporated within the uranium alteration phases and suggest that ionic substitution and/or adsorption to the uranyl phases can play a key role in the limiting the mobility of REE (and by analogy, actinide elements) in a nuclear waste repository.     

Hypothetical integration of a fast reactor to the fuel cycle in Slovakia

The inspiration for dealing with the topic of fuel cycle back-end was attendance at a European project called RED-IMPACT – Impact of Partitioning Transmutation and Waste Reduction Technologies. This paper includes an image how to re-use energetic potential of stored spent fuel and at the same time how to effectively reduce spent fuel and radioactive waste volumes aimed for deep repositories. The first part is based on the analysis of Pu and minor actinides (MA) content in actual VVER-440 spent fuel stored in Slovakia. The next parts present the hypothetical possibilities of reprocessing and Pu re-use in a fast reactor under Slovak conditions. For the hypothetical transmutation of heavy nuclides (Pu and MA) contained in Slovak spent fuel a SUPERPHENIX (SPX) fast reactor with increased power was chosen because a fast nuclear reactor cooled by sodium belongs to the group of Generation IV reactor systems. This article deals with the analysis of power production and fuel cycle indicators. The indicators of the SPX calculation model were compared with the results of the VVER-440 spent fuel with the initial fuel enrichment of 4.25% U-235 + 3.35% Gd₂O₃. The created SPX model in the spectral computer code HELIOS 1.10 consists of a fissile (fuel) and a fertile part (blanket). All kinds of calculations were performed by the computer code HELIOS 1.10. This study also exposes the HELIOS modelling and simulating borders.     

Operational experience of on-site transport of spent nuclear fuel in Korea

Abstract

With the rapid development of the nuclear power programme in Korea, the amount of accumulated spent nuclear fuel has inevitably increased year by year. The spent nuclear fuel is being stored in on-site storage pools at the nuclear power plants. As the current storage capacity for spent nuclear fuel is insufficient, at-reactor storage is being expanded at each site with regard to optimisation of technical and economic factors. On-site transport between neighbouring reactors has been necessary to secure sufficient storage capacity for pressurised water reactor spent nuclear fuel assemblies. A complete on-site transport system has been developed, and so far more than 800 spent nuclear fuel assemblies have been transported using two kinds of transport cask.