Understanding long-term corrosion of Alloy 22 container in the potential Yucca Mountain repository for high-level nuclear waste disposal

Alloy 22 (Ni–22Cr–13Mo–3W–4Fe) is the candidate material for the waste package outer container in a potential geologic repository for high-level nuclear waste disposal at Yucca Mountain, Nevada. This alloy exhibits very low corrosion rates in the absence of environmental conditions promoting crevice corrosion. However, there are uncertainties regarding Alloy 22’s corrosion performance when general corrosion rates and susceptibility to crevice corrosion are extrapolated to a geological time period (e.g. 10⁵ years). This paper presents an analysis of available literature information relevant to the long-term extrapolation of general corrosion processes and the crevice corrosion behavior of Alloy 22, under potential repository environments. For assessment of general corrosion rates, potential degradation processes causing the loss of the long-term persistence of passive film formed are considered. For crevice corrosion, induction time, and the extent of susceptibility and opening area, are considered. Disclaimer: The US Nuclear Regulatory Commission (NRC) staff views expressed herein are preliminary and do not constitute a final judgment or determination of the matters addressed nor of the acceptability of a license application for a geologic repository at Yucca Mountain. The paper describes work performed by the Center for Nuclear Waste Regulatory Analyses (CNWRA) for NRC under Contract Number NRC-02-02-012. The activities reported here were performed by CNWRA on behalf of the NRC office of Nuclear Material Safety and Safeguards, Division of High Level Waste Repository Safety. This paper is an independent product of the CNWRA and does not necessarily reflect the view or regulatory position of the NRC.     

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.     

基于WebGIS的北山高放废物处置地学信息系统的研究

讨论了WebGIS技术,选择GeoServer和PostgreSQL等开源软件来搭建WebGIS系统,采用Web浏览器、WebGIS服务器和数据服务器三层构架。将使用ASP.NET技术开发基于WebGIS的北山高放废物处置地学信息系统,使用户通过Internet,以图形化的方式方便地浏览、操作和分析北山地区各种地学信息,从而实现高效的地学信息共享,为北山地学信息的管理决策提供科学依据。     

Failure model and Monte Carlo simulations for titanium (grade-7) drip shields under Yucca Mountain repository conditions

A failure model was developed for the titanium alloy drip shield proposed for the Yucca Mountain nuclear waste repository. The degradation modes considered are hydrogen-induced cracking (HIC) and general aqueous corrosion, processes which are inextricably linked. Failure by HIC is controlled by the environment, the corrosion rate, the material properties, and the hydrogen absorption efficiency which is assumed to decrease parabolically with time. This model includes both oxygen and water reduction coupled to corrosion, and allows for the release of the absorbed hydrogen as the alloy containing the hydrogen converts to oxide. Monte Carlo simulations were employed to predict drip shield lifetimes, and to investigate the effects of the hydrogen absorption efficiency, the critical HIC concentration, the corrosion rate, and the fraction of corrosion supported by water reduction, on the susceptibility of the material to HIC.     

Second generation waste package design and storage concept for the Yucca Mountain Repository

The reference waste package design and operating mode to be used in the Yucca Mountain Repository is reviewed. An alternate (second generation) operating concept and waste package design is proposed to reduce the risk of localized corrosion of waste packages and to reduce repository costs. The second generation waste package design and storage concept is proposed for implementation after the initial licensing and operation of the reference repository design. Implementation of the second generation concept at Yucca Mountain would follow regulatory processes analogous to those used successfully to extend the design life and uprate the power of commercial light water nuclear reactors in the United States. The second generation concept utilizes the benefits of hot dry storage to minimize the potential for localized corrosion of the waste package by liquid electrolytes. The second generation concept permits major reductions in repository costs by increasing the number of fuel assemblies stored in each waste package, by eliminating the need for titanium drip shields and by fabricating the outer container from corrosion resistant low alloy carbon steel.     

Migration processes in a model deep repository for the disposal of intermediate-level nuclear waste

It is difficult to conceive of radionuclides escaping from a repository by any means other than migration in groundwater. Simple models of the repository are constructed and various migration processes are identified and assessed, according to the flow speed of water through the repository. Diffusion in static water and advection in fast flows are considered separately initially, but later we examine the effect of slow flows in which both these processes contribute to the removal of radionuclides. Concentration profiles across the repository, fluxes of nuclides and total losses are obtained from the analysis. We investigate the time scales necessary for the steady state to be achieved in the repository and conclude that flow speed is roughly inversely proportional to this time scale, i.e. faster flows establish a steady state sooner than slow ones. We also assess the sensitivity of the results to the physical properties of the components of the repository.     

高放废物处置库场址北山预选区新构造运动分区与中、强地震分布特征

根据高放废物处置库场址北山预选区的新构造运动特征,将其分为祁连山强烈隆升区、河西走廊差异沉降区、安西差异沉降区和北山弱隆升区。预选区内中、强地震集中分布于祁连山强烈隆升区和河西走廊差异沉降区,其中,5.0级以上地震的发生均与区内主要大型断裂带的活动有关。中、强地震的分布与新构造运动的活跃程度相关。新生代以来,青藏高原的NE向扩展、挤压和缩短是形成以上新构造运动格局及地震分布的动力来源。     

Efficiency analyses of the CANDU spent fuel repository using modified disposal canisters for a deep geological disposal system design

Deep geological disposal concept is considered to be the most preferable for isolating high-level radioactive waste (HLW), including nuclear spent fuels, from the biosphere in a safe manner. The purpose of deep geological disposal of HLW is to isolate radioactive waste and to inhibit its release of for a long time, so that its toxicity does not affect the human beings and the biosphere. One of the most important requirements of HLW repository design for a deep geological disposal system is to keep the buffer temperature below 100 °C in order to maintain the integrity of the engineered barrier system. In this study, a reference disposal concept for spent nuclear fuels in Korea has been reviewed, and based on this concept, efficient alternative concepts that consider modified CANDU spent fuels disposal canister, were developed. To meet the thermal requirement of the disposal system, the spacing of the disposal tunnels and that of the disposal pits for each alternative concept, were drawn following heat transfer analyses. From the result of the thermal analyses, the disposal efficiency of the alternative concepts was reviewed and the most effective concept suggested. The results of these analyses can be used for a deep geological repository design and detailed analyses, based on exact site characteristics data, will reduce the uncertainty of the results.     

高放废物地质处置元数据设计与编辑模块开发

在对元数据的含义、作用、组成及对高放领域研究的数据内容和特点分析的基础上,对元数据编辑模块的设计进行了详细描述,并在VS2005平台下应用C#语言进行了模块的实际开发,最终实现了便于进行数据共享的XML格式存储功能。元数据编辑模块的应用将促进我国高放废物地质处置信息化工作的进一步深入。     

Review of waste glass corrosion and associated radionuclide release as a part of safety assesment of entire disposal system

Current knowledge on high-level nuclear waste glass corrosion is summarized, and remaining problems are discussed for meaningful predictions of the glass corrosion and associated radionuclide release as a part of safety assessment of entire disposal system. In recent years, much progress has been made in understanding the mechanism of waste glass corrosion in aqueous environments. Glass corrosion models based on the mechanism have been developed for predicting the long-term glass performance, and they are incorporated as part of radionuclide source term in safety assessments of the disposal system. However, these results have not yet allowed meaningful predictions for the long-term release of individual radionuclides from the glass in repository environments, because mechanism of the long-term glass corrosion has not been fully understood and solubilities of actinoids and fission products under disposal conditions are rather uncertain. In addition, the most serious problem is that the effects of various reactions and interactions occurring in the engineered barrier system, such as corrosion of overpack, alteration of backfill and chemical interactions of the released glass constituents with them have not been fully coupled with the glass performance. These reactions may be dominant processes controlling the glass corrosion and associated radionuclide release for the long-term. For the meaningful predictions, we must evaluate the waste glass performance in combination with the effects of various reactions and interactions occurring in the engineered barrier system on the basis of fully understanding of the chemical and geochemical mechanisms.     

高放废物处置库近场裂隙水流-传热-处置室间距相互作用的三维数值分析

为研究高放废物地质处置库近场裂隙水流-传热-处置室间距的相互作用机理,采用3DEC软件计算裂隙水流-传热-处置室间距相互作用对处置库近场温度分布影响。结果表明:(1)在处置室间距相同条件下,流动的裂隙水显著改变了处置库近场温度场,使岩体温度降低,缩短模型达到稳态所需要的时间。(2)处置室间距增大,温度叠加效应减弱,处置库近场温度越低,并且废物罐表面膨润土温度越低,裂隙出水口水温越低,模型达到稳态所需要的时间越短。(3)水平和垂直裂隙水流共同传热使处置库近场裂隙水流下游区域温度显著高于裂隙水流上游区域。(4)处置室间距为6 m和8 m时,水平裂隙出水口水温高于垂直裂隙,处置室间距为10 m时,水平裂隙出水口水温低于垂直裂隙。     

Analyses of disposal efficiency based on nuclear spent fuel cooling time and disposal tunnel/pit spacing for the design of a geological repository

The purpose of deep geological disposal of high-level radioactive waste (HLW) including nuclear spent fuels is to isolate and to inhibit the release of radioactive material for a long time so that its toxicity does not affect the biosphere. The main requirement for the HLW repository design is to keep the buffer temperature below 100 °C in order to maintain the integrity of the engineered barrier system. The cooling time of the spent fuels discharged from nuclear power plants is the key consideration factor for the efficiency and economic feasibility of such a repository. We analyze the spacing of the disposal tunnels and pits, the disposal area and the uranium density for the deep geological repository layout to satisfy the thermal requirement of the disposal system. To do this, thermal stability analyses of a disposal system have been performed using varying spent fuel cooling times and spacing of the disposal tunnels and pits. The results show that the time to reach the maximum temperature within the design limit of the temperature in the disposal site is likely to be shortened as the cooling time of the spent fuel becomes shorter. Also it seems that controlling the disposal pit spacing is considered more advantageous than controlling the disposal tunnel spacing to meet the allowable thermal criteria in the repository from thermal and economical points of view. The results of these analyses can be used for a deep geological repository design and detailed analyses with exact site characteristics data will reduce the uncertainty of the results.     

放射性废物处置安全全过程系统分析及有关问题的探讨

本文根据IAEA第SSG-23号安全导则,对放射性废物处置安全全过程系统分析及其在英国、法国、美国、芬兰、瑞士、瑞典等国家放射性废物处置中的应用情况作了概括性的介绍。结合我国放射性废物处置管理的现状,对我国处置场的环境影响评价和安全分析中存在的主要问题进行了探讨。为促进我国安全全过程系统分析工作的全面开展,建议加快制定相关标准,将“安全全过程系统分析”作为我国放射性废物处置的许可条件,在景象开发、长期演变、坚稳性、不确定性、管理系统等方面加强研究。     

Concept of optimization for geologic disposal combined with P&B/T (partitioning, and burning and/or transmutation) treatment in self-completed fuel cycle

Self-Completed Fuel Cycle combined with multiple-recycle P&B/T (partitioning, and burning and/or transmutation) treatment has the potential merits for improving the present geologic disposal scenario, and is discussed herein based on three plus one criteria for partitioning. The number reduction of GSC (glass solidified canister, i.e. waste package) is important concept to decrease the cost of geologic disposal. The cost needed for P&B/T treatment can be saved by the number reduction of GSC than the current disposal scenario.

The total cost for P&B/T treatment with thermal and fast B/T reactors can be minimized by the variation of recycle period, and by the adjustment of out-core optimization and in-core optimization. The partitioning can be done by three plus one criteria, that is (1) the selection of long-lived radio-nuclide with high Hazard Index at 1,000 years after the fuel discharge, (2) the selection of radio-nuclide with high mobility through deep geologic media, and (3) the grouping of MA (minor actinide) and LLFP (long-lived fission product) selected by the items (1) and (2), based on B/T characteristics and decay acceleration ratio of nuclear reactions.     

Development of a web-based sorption database (KAERI-SDB) and application to the safety assessment of a radioactive waste disposal

Sorption plays a key role in a retardation of radionuclide migration in various geological environments. Hence sorption of radionuclides onto geological media is one of the important factors for the safety assessment of radioactive waste disposal. A web-based radionuclide sorption database program named KAERI-SDB has been developed to provide a database for the sorption of radionuclides onto geological media at various geochemical conditions. The KAERI-SDB is designed to determine the distribution coefficient (K_d) of a radionuclide and evaluate sorption properties by easily accessing an internet web-site (http://sdb.kaeri.re.kr). The KAERI-SDB provides a useful output and search result as a scatter plot chart or an index chart. The KAEI-SDB was designed to show the search results in a statistical way by representing the mean K_d value at 95% of confidence as a function of major geochemical indices. Several case studies were carried out to demonstrate the applicability of the KAERI-SDB and the result showed a successful applicability of the KAERI-SDB to various radionuclide sorption cases.     

Multiple lines of evidence for performance of the canister and waste form in long-term nuclear waste disposal: Reviews

The time scales required for nuclear waste disposal are very large compared with those for other engineering endeavors. Because of this, there are many uncertainties associated with the quantitative performance assessment of canisters containing high-level radioactive waste in a waste form. Multiple lines of evidence can be helpful in building confidence in the long-term behavior (corrosion and dissolution) of the canister and waste form. These lines of evidence are derived from long-term supports and probabilistic models and developed based on shorter term tests, bounding and conservative approaches, and available observations on natural analogs. This paper presents the progress made for important lines of evidence considered in quantitatively assessing radionuclide release behavior from canisters and waste forms. This paper considers risk-significant issues for canisters and waste forms (i.e., risk informed approach) in the probabilistic performance assessment of the disposal system which has also other components such as geology and hydrology.     

U.S. accelerator-driven transmutation of waste (ATW) program objectives, and technology development efforts

The U.S. Program to evaluate accelerator-driven systems for transmuting problematic, long-lived nuclear waste stream components was initiated during fiscal year 2000, based largely on the Accelerator-driven Transmutation of Waste (ATW) Technology Development Road Map developed during 1999. The Road Map (DOE/RW-0519) effort provided a long-range plan, involving technology development, demonstration, and deployment, as well as a recommended initial effort to evaluate the technology options for five or six years. This paper summarizes the ATW Research and Development Plan currently in draft form. Technology Readiness Levels (TRLs), which are based in part on a system used by the U.S. National Aeronautics and Space Administration in determining levels of flight readiness, was developed for use in assessing and advancing technologies relevant to waste transmutation. Based on TRLs and other considerations, the Program is screening technology options and prioritizing the long-term research and development effort. A top-level schedule illustrates the efforts planned to advance the important technology options in preparation for integrated system tests.     

聚变材料数据库平台FUMDS的设计与发展

聚变材料数据库平台(FUMDS,FUsion Materials Database System)收集了聚变材料研究相关的各种材料的数据,运用现代计算机技术和数据分析方法,对材料数据进行系统的管理和分析,为聚变堆设计以及聚变材料的研究和工程制备提供数据支持.     

The effect of the evolution of environmental conditions on the corrosion evolutionary path in a repository for spent fuel and high-level waste in Opalinus Clay

The corrosion evolutionary path (CEP) defines the time-dependent corrosion behaviour of canisters in a deep geologic repository. In turn, the CEP is largely determined by the evolution of the environmental conditions in the near- and far-fields. The evolution of environmental conditions in a repository for spent fuel and high-level waste in Opalinus Clay is described along with the impact on the corrosion behaviour of the canisters.