Development of compartment models with Markov-chain processes for radionuclide transport in repository region

This paper presents a new radionuclide transport model for performance assessment and design of a geologic repository for high-level radioactive waste. The model uses compartmentalization of a model space and a Markov-chain process to describe the transport. The model space is divided into an array of compartments, among which a transition probability matrix describes radionuclide transport. While similar to the finite-difference method, it has several advantages such as flexibility to include various types of transport processes and reactions due to probabilistic interpretation, and higher-order accuracy resulting from direct formulation in a discrete-time frame.We demonstrated application of this model with a hypothetical repository in porous rock formation. First we calculated a three-dimensional steady-state heterogeneous groundwater flow field numerically by the finite-element method. The transition probability matrix was constructed based on the flow field and hydraulic dispersion coefficient. The present approach has been found to be effective in modeling radionuclide transport at a repository scale while taking into account the effects of change in hydraulic properties on the repository performance. Numerical exploration results indicate that engineered barrier configuration and material degradation have substantial effects on radionuclide release from the repository.     

The effectiveness of full actinide recycle as a nuclear waste management strategy when implemented over a limited timeframe – Part II: Thorium fuel cycle

Full recycling of transuranic (TRU) isotopes can in theory lead to a reduction in repository radiotoxicity to reference levels in as little as ∼500 years provided reprocessing and fuel fabrication losses are limited. However, over a limited timeframe, the radiotoxicity of the ‘final’ core can dominate over reprocessing losses, leading to a much lower reduction in radiotoxicity compared to that achievable at equilibrium. In Part I of this paper, TRU recycle over up to 5 generations of light water reactors (LWRs) or sodium-cooled fast reactors (SFRs) is considered for uranium (U) fuel cycles. With full actinide recycling, at least 6 generations of SFRs are required in a gradual phase-out of nuclear power to achieve transmutation performance approaching the theoretical equilibrium performance. U-fuelled SFRs operating a break-even fuel cycle are not particularly effective at reducing repository radiotoxicity as the final core load dominates over a very long timeframe. In this paper, the analysis is extended to the thorium (Th) fuel cycle. Closed Th-based fuel cycles are well known to have lower equilibrium radiotoxicity than U-based fuel cycles but the time taken to reach equilibrium is generally very long. Th burner fuel cycles with SFRs are found to result in very similar radiotoxicity to U burner fuel cycles with SFRs for one less generation of reactors, provided that protactinium (Pa) is recycled. Th-fuelled reduced-moderation boiling water reactors (RBWRs) are also considered, but for burner fuel cycles their performance is substantially worse, with the waste taking ∼3–5 times longer to decay to the reference level than for Th-fuelled SFRs with the same number of generations. Th break-even fuel cycles require ∼3 generations of operation before their waste radiotoxicity benefits result in decay to the reference level in ∼1000 years. While this is a very long timeframe, it is roughly half that required for waste from the Th or U burner fuel cycle to decay to the reference level, and less than a tenth that required for the U break-even fuel cycle. The improved performance over burner fuel cycles is due to a more substantial contribution of energy generated by ²³³U leading to lower radiotoxicity per unit energy generation. To some extent this an argument based on how the radiotoxicity is normalised: operating a break-even fuel cycle rather than phasing out nuclear power using a burner fuel cycle results in higher repository radiotoxicity in absolute terms. The advantage of Th break-even fuel cycles is also contingent on recycling Pa, and reprocessing losses are significant also for a small number of generations due to the need to effectively burn down the TRU. The integrated decay heat over the scenario timeframe is almost twice as high for a break-even Th fuel cycle than a break-even U fuel cycle when using SFRs, as a result of much higher ⁹⁰Sr production, which subsequently decays into ⁹⁰Y. The peak decay heat is comparable. As decay heat at vitrification and repository decay heat affect repository sizing, this may weaken the argument for the Th cycle.     

城市放射性废物库设计探讨

针对江苏省放射性废物的特点,提出城市放射性废物库库型、吊装工具、防洪抗震、安全防御及污染防治等方面的设计方案,既保证江苏省放射性废物和废旧放射源能得到安全收贮,又确保库区环境不受污染,促进江苏省核技术利用企业的发展。     

高放废物地质处置黏土岩处置库围岩研究现状

世界上很多国家都对处置库的可能围岩进行了详细研究。通过对比,认为花岗岩、黏土岩、岩盐比较适合作为处置库围岩,而黏土岩由于具有自封闭性、渗透率低等其他岩石类型不可比拟的优点,因而将黏土岩作为高放废物地质处置库围岩越来越受到各国的关注。文章同时介绍了瑞士、法国、比利时等国家在黏土岩中所进行的大量研究,均认为在黏土岩中处置高放废物和乏燃料是安全的。文章还对黏土岩处置库概念设计、黏土岩处置库围岩地下实验室研究,以及我国开展黏土岩处置库研究的意义等进行了综述。     

Testing of the Sealing Arrangements for the Nirex Re-Usable Shielded Transport Containers

Abstract

As part of its responsibility for the development of a deep repository for intermediate and low level radioactive waste, UK Nirex Ltd is developing a range of Type B re-usable shielded transport containers (RSTCs). A testing programme has been carried out on two alternative concepts for the RSTC sealing arrangements over the temperature range ?40^°C to 200^°C. For each sealing system, a test rig was developed to measure the performance under simulated normal and accident conditions of transport. The elastomer O-rings used for some of the tests had been irradiated to the maximum dose they might receive in normal transport. The performance of both sealing systems was good and it is concluded that either concept would meet the specified leakage criteria over the full temperature range under both normal and accident conditions of transport. However, further testing is required to confirm the performance of Concept N under accident conditions.     

Monte Carlo estimation of radionuclide release at a repository scale

Prediction of radionuclide release is a central issue in the performance assessment of nuclear waste repositories. This requires modeling of the radionuclide migration processes through the repository barriers, accounting for the related uncertainties. The present paper illustrates a Monte Carlo simulation-based compartment model in which detailed, local-scale modeling feeds a global-scale analysis of the repository, at reasonable computational expenses. An application to a realistic case study is presented to verify the feasibility of the approach.     

缓冲回填材料膨润土胶体的提取与分析

我国拟建造的高放废物地质处置库计划使用膨润土作为缓冲回填材料。有效提取膨润土在处置库地下水中形成的胶体,检测其各项物理化学参数,对理解胶体对关键放射性核素在处置库近场的吸附、扩散和迁移行为的影响具有重要意义。本工作分析了原状高庙子膨润土的矿物组成及其元素含量。对原状高庙子膨润土胶体预处理、提取方法进行了优化,建立了稳定、可靠获取膨润土胶体的实验方法。结果表明：超声振荡或沸热分散后再离心分离的方法不仅能在较短的时间内获得尺寸在100 nm左右的膨润土胶体,且有较好的单体分散性,此外,不同批次胶体样品的zeta电位均大于-60 mV,显示其良好的稳定性。能量色散X射线光谱分析结果表明,膨润土胶体主要成分为SiO₂。     

The UK rock characterization programme

The objective of the Nirex Science Programme is to evaluate the post-closure safety performance of a deep repository for intermediate level wastes (both long and short lived) and some low level wastes at a site near Sellafield in West Cumbria. The extensive surface-based site characterization programme, complemented by research and assessment studies, has established the potential suitability of the Sellafield site for the deep repository. However, a stage of underground investigations, the Rock Characterization Facility, is needed in order to establish the high level of confidence necessary to underpin a decision on whether to propose development of a repository, and subsequently a decision by the regulators to authorize commencement of disposal operations in the repository when it is constructed. The Rock Characterization Facility will be developed in three phases, providing a scientific programme lasting around ten years. It will address three key areas of uncertainty as follows: (1) groundwater flow and radionuclide transport; (2) natural and induced changes to the geological barrier; (3) design and construction of the repository. The nature of these uncertainties, the way in which underground investigations enable us to resolve them, and the design of the Rock Characterization Facility and its experimental programme to deliver the required information to resolve these uncertainties are described in this paper.     

Experimental determination and numerical simulation of the permeability of cementitious materials

Safety analysis of a repository requires a detailed numerical study of the coupled transport processes of gas and water in the repository. Experimental studies were carried out to measure the transport parameters of different types of normal concrete. The capillary pressure curve, the gas and the water permeabilities were determined in separate experiments. Especially the influence of the water content of the samples was under investigation. Additionally coupled gas and water flow experiments were carried out and numerically simulated with the measured transport parameters. With the relative permeability curve and the experimental determined pore size distribution it is possible to describe the coupled transport of gas and water through specimens.     

Evaluation of nuclide release scenarios for a hypothetical LILW repository

A program for the safety assessment and performance evaluation of a low- and intermediate-level radioactive waste (LILW) repository system has been developed. Utilizing GoldSim (2006), the program evaluates nuclide release and transport into the geosphere and biosphere under various disruptive natural and manmade events and scenarios that can occur after a waste package failure. We envisaged and illustrated these events and scenarios as occurring after the closure of a hypothetical LILW repository, and they included the degradation of various manmade barriers, pumping well drilling, and natural disruptions such as the sudden formation of a preferential flow pathway in the far-field area of the repository. Possible enhancement of nuclide transport facilitated by colloids or chelating agents is also dealt with. We used the newly-developed GoldSim template program, which is capable of various nuclide release scenarios and is greatly suited for simulating a potential repository given the geological circumstances in Korea, to create the detailed source-term and near-field release scheme, various nuclide transport modes in the far-field geosphere area, and the biosphere transfer. Even though all parameter values applied to the hypothetical repository were assumed, the illustrative results, particularly the probabilistic calculations and sensitivity studies, may be informative under various scenarios. (GoldSim, LILW, Nuclide transport, Safety assessment, Scenario).     

辐射和热对高放废物地质处置库缓冲材料性能演化的影响

辐射和释热是高放废物的内在特性,辐射和热对膨润土性能的影响对于高放废物地质处置库安全评价至关重要。本文对国外近年来有关辐射和热对工程屏障材料性能的影响研究成果进行了初步的调研,其中有关热对工程屏障材料性能影响的调研主要集中关注了在较高温度(100 ℃以上)时工程屏障材料性能的变化。     

高放废物处置库选址中的水文地质特性评价

根据国内外研究经验,综述高放废物地质处置库选址不同研究阶段水文地质特性评价的方法和作用,并将评价方法分为区域、地区和场址3个阶段水文地质调查来讨论。同时,介绍了我国高放废物地质处置场地水文地质特性评价研究现状。     

Integrated decay heat load method to analyze repository capacity impact of a fuel cycle

Assessing the needs for repository capacity from nuclear waste disposal is essential for fuel cycle development or repository development planning. As the repository capacity is mainly constrained by thermal design limits on the repository rocks, a detailed mountain-scale heat transfer calculation is needed for repository capacity impact analysis. In this paper, a simplified repository capacity impact analysis method is proposed as an alternative to performing repository scale heat transfer analysis. The method is based on the use of integrated decay heat load (IDHL) limits. The derived integrated decay heat loads were found to appropriately represent the drift wall temperature limit (200 °C) and the midway between adjacent drifts temperature limit (96 °C) under the high temperature operating mode as long as the wastes are uniformly loaded into the repository. Results indicated that the long-term integrated decay heat load (IDHL_L) and the short-term integrated decay heat load (IDHL_S) can be effectively used to represent the repository capacity impact for SNFs and HLWs, respectively. Comparisons indicated good agreement between the proposed IDHL method and the repository heat transfer analysis-based approach.     

Three-dimensional elastic recoil detection analysis of hydrogen in materials for use in the Yucca Mountain nuclear waste repository

An original system has been developed capable of performing three-dimensional Energy Recoil Detection Analysis (ERDA) of hydrogen in materials with scanned and finely focused heavy-ion beams. The technique is being used at the Lawrence Livermore National Laborotory Multi-user Tandem Laboratory to measure the hydrogen content in materials under consideration for use in the Yucca Mountain nuclear waste repository. From the measurement of hydrogen concentration profiles we can extrapolate reaction rates. A critical problem is the rate of dissolution of the glass being used. The HI-ERDA (Heavy Ion-ERDA) technique can provide this information which is needed in order to predict the overall rate of nuclear waste glass degradation in a waste repository. The technique is calibrated using a silicon wafer implanted with a known concentration of hydrogen. The sample is illuminated by a ³⁵Cl ion beam that is micro-scanned across the sample. From these measurements we reconstruct three dimensional profiles of hydrogen content which can then be used to obtain spatially-resolved hydrogen depth profiles.     

An evaluation of nuclide release from a trench-type LILW repository

A simple and effective model and a GoldSim (GoldSim, 2006) template program, by which a probabilistic safety assessment of a conceptual trench-type repository for low- and intermediate level radioactive waste (LILW) disposal can be carried out under various nuclide release scenarios, have been developed. To quantify the exposure dose rates due to nuclide release from the trench and transport through the various pathways possible in the near- and far-fields of the repository system under a base case and some alternative scenarios, illustrative evaluations for a comparison among the scenarios as well as a sensitivity of shortcut pathways generated due to earthquakes on the nuclide transport are made and demonstrated. To this end, by changing the conservative base case nuclide release scenario under which all portions of the cap of the trench are failed unconditionally and immediately after a closure of the repository, a total of four other alternative scenarios were separately evaluated for the total exposure dose rates to the farming exposure group and then compared to the base case results. Among them, an earthquake scenario shows a dominant behavior almost throughout the whole time span. To see the influence of all shortcut travel pathways that are assumed to be newly generated by this earthquake scenario, their sensitivities to the exposure dose rates to the farming exposure group were also made and compared to each other.     

化学种态分析软件CHEMSPEC最新进展

我国高放废物地质处置研发工作已经步入建造地下实验室阶段。地下实验室建造安全评价和未来的处置库性能评价中均需要关键放射性核素在相应深部地质条件下的扩散和迁移参数,而关键核素的扩散和迁移参数与核素在相应水岩体系中的化学种态密切相关。为满足我国核设施退役治理工作的需要,尤其是我国高放废物地质处置相关安全评价的需要,北京大学核环境化学课题组于2008年开始编写具有完全著作权的化学种态分析软件CHEMSPEC。经过多次修改和完善,目前已经具备了较好的计算功能。本文介绍该软件在表面配合模型、数据库补充和程序优化方面的最新进展,以实例形式介绍该软件的新性能,以期为我国相关实验研究者使用该软件提供参考。     

花岗岩中高放废物地质处置库系统分析方法研究

首先以甘肃北山预选区花岗岩场址为例,提出该场址中高放废物地质处置库概念设计和结构设计,然后以系统分析方法论为基础,描述处置库的系统功能、结构、环境及其演化过程.并以模拟软件GoldSim为工具,建立该处置库演化过程的计算机模型,最后以该计算机模型为模拟实验平台,模拟处置库中辐射毒性时空分布,分析模型中的参数灵敏度,优化设计参数,并预测和评价处置库性能.其研究成果可为合理配置资源和有效协调各研究项目之间关系提供技术支持.     

Towards a Swedish repository for spent fuel

Nuclear power is producing electricity for the benefit of society but is also leaving radioactive residues behind. It is our responsibility to handle these residues in a safe and proper manner. The development of a system for handling spent fuel from nuclear power plants has proceeded in steps. The same is true for the actual construction of facilities and will continue to be the case for the final repository for spent fuel and other types of long-lived wastes. The primary objective in constructing the repository will be to isolate and contain the radioactive waste. In case the isolation fails for some reason the multibarrier system should retain and retard the radionuclides that might come into contact with the groundwater. A repository is now planned to be built in two steps where the first step will include deposition of about 400 canisters with spent fuel. This first step should be finished in about 20 years from now and be followed by an extensive evaluation of the results from not only this particular step but also from the development of alternative routes before deciding on how to proceed. A special facility to encapsulate the spent fuel is also required. Such an encapsulation plant is proposed to be constructed as an extension of the existing interim storage CLAB. Finding a site for the repository is a critical issue in the implementation of any repository. The siting process started a few years ago and made some progress but is by no means yet completed. It will go on at least into the early part of the next decade. When the present nuclear power plants begin to be due for retirement there should also be some facilities in place to take permanent care of the long-lived radioactive residues. Progress in siting will be a prerequisite for success in our responsibility to make progress towards a safe permanent solution of the waste issue.     

Suction and water uptake in unsaturated compacted bentonite

Suction is an important process dominating water movement in unsaturated compacted bentonite. Suction measurements were carried out to investigate the characteristics of suction in unsaturated compacted bentonite for the buffer of a HLW repository. The suction values decreased with increasing the water content at a given dry density, and it revealed higher value at the higher dry density of compacted bentonite. The suction variation with an increase of temperature was negligible for the water content up to 17%. For the dry density of 1600 kg/m³, the suction value in the vertical measurement was a little lower than that in the horizontal one due to the microstructural anisotropy of compacted bentonite. The modeling of water uptake in unsaturated compacted bentonite is allowed to predict the re-saturation for the buffer of a HLW repository. To this end, water uptake tests were conducted to compare their results with those calculated using the coupled hydro-mechanical model of a computer code ABAQUS. Both results were in reasonably good agreement with each other, which suggests that the model can be used to predict the re-saturation of the bentonitic buffer in a HLW repository.