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.     

Review of the development of the transportation,aging, and disposal (TAD) waste disposal system for the proposed Yucca Mountain geologic repository

The U.S. Department of Energy (DOE) began studying Yucca Mountain in 1978 to determine whether it would be suitable for the nation’s first long-tem geologic repository for over 70,000 metric tons of spent (or used) nuclear fuel and high-level radioactive waste. The purpose of the continuing Yucca Mountain study, or project, is to comply with the Nuclear Waste Policy Act of 1982 as amended in 1987 and develop a national disposal site for spent nuclear fuel and high-level radioactive waste disposal. In 2005, DOE shifted the design of the proposed repository from a concept of unloading spent nuclear fuel from transportation canisters and loading into disposal canisters (which required a great deal of handling radioactive material at the repository site) to a “clean” facility, unveiling the transportation, aging, and disposal (TAD) canister system. The TAD waste system consists of a canister loaded with commercial spent nuclear fuel.This review paper provides a comprehensive review on the status of TAD, technical and licensing requirements, the work that has been done so far, and the challenges and issues that must be addressed before TAD can be successfully implemented. Though the future of the Yucca Mountain project is bleak at this point, the progress that has come in the field of TAD will be one of its lasting legacies.     

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.     

今日的尤卡山计划和多重屏障系统的新概念

主要论述了美国尤卡山项目的最新进展和比利时工程屏障研究的新进展。在尤卡山项目新进展方面介绍了尤卡山项目修改的近期计划,以及两个环境评价补充报告和两位民主党参议员参加2008年美国总统竞选时对尤卡山项目的态度。还介绍了比利时为加强高放废物地质处置的安全性而开发的多重屏障的新概念。     

Interim storage of power reactor spent nuclear fuel (SNF) and its potential application to SNF separations and closed fuel cycle

Interim, centralized, engineered (dry cask) storage facilities for USA light water power reactor spent nuclear fuel (SNF) should be implemented to complement and to offer much needed flexibility while the Nuclear Regulatory Commission is funded to complete its evaluation of the Yucca Mountain License and to subject it to public hearings. The interim sites should use the credo reproduced in Table 1 [Bunn, M., 2001. Interim Storage of Spent Nuclear Fuel. Harvard University and University of Tokyo] and involve both the industry and government. The sites will help settle the 50 pending lawsuits against the government and the $11 billion of potential additional liabilities for SNF delay damages if Yucca Mountain does not being operation in 2020 [DOE, 2008a. Report to Congress on the Demonstration of the Interim Storage of Spent Nuclear Fuel from Decommissioned Nuclear Power Stations (December)].Under the developing consensus to proceed with closed fuel cycles, it will be necessary to develop SNF separation facilities with stringent requirements upon separation processes and upon generation of only highly resistant waste forms. The location of such facilities at the interim storage sites would offer great benefits to those sites and assure their long term viability by returning them to their original status.The switch from once-through to closed fuel cycle will require extensive time and development work as illustrated in “The Path to Sustainable Nuclear Energy” [DOE, 2005. The Path to Sustainable Nuclear Energy. Basic and Applied Research Opportunities for Advanced Fuel Cycles. DOE (September)]. A carefully crafted long term program, funded for at least 5 years, managed by a strong joint government–industry team, and subjected to regular independent reviews should be considered to assure the program stability and success.The new uncertainty about Yucca Mountain role raises two key issues: (a) what to do with the weapons and other high level government wastes committed to be moved to Yucca Mountain by specified dates? And (b) can the $13.6 billion invested at Yucca Mountain be salvaged if the NRC approves the license submittal and the opposition relents after contentious hearings? Or will it take contingent actions, or, a switch to a partial closed fuel cycle with its reduced risks and earlier timing of their peak risk value? Only time will tell if any of these alternates will be acceptable but, they all reinforce the need for interim storage for commercial SNF.If the decision is to go to a new repository one wonders whether the time has not come to change the safety evaluation process for geological repositories by characterizing two to three sites and subjecting them to an arbitrary release of the fission products in the HLW to be stored and considering the forms of some of the HLW to reduce their peak risks. It would allow the proper choice to be made among the selected sites and to have a basis for convincing the local committee to accept the repository location. It may even decide whether the CONFU fuel assembly [MIT, 2006. Implications of alternative strategies for transition to sustainable fuel cycles. Nucl. Sci. Eng., 154 (September)] for pressurized water reactors can be accommodated in a once-through fuel cycle as suggested by Levy [Levy, S., 2008. Yucca backup plan. Nucl. Eng. Int., 24–28]. A similar configuration is possible in boiling water reactors.     

Probabilistic methodology to estimate environmental conditions for localized corrosion and stress corrosion cracking of Alloy 22 in a high-level radioactive waste repository setting

The US Department of Energy (DOE) has indicated that it may use Alloy 22 (Ni-22Cr-13Mo-4Fe-3W) as the waste package outer container material for the potential high-level waste repository at Yucca Mountain, Nevada. This alloy could be susceptible to localized corrosion, in the form of crevice corrosion, and stress corrosion cracking if environmental conditions and material requirements (e.g., existence of crevices or high enough tensile stresses) are met. An approach is proposed to assess the likelihood of environmental conditions capable of inducing crevice corrosion or stress corrosion cracking in Alloy 22. The approach is based on thermodynamic simulations of evaporation of porewaters and published equations to compute corrosion potential and critical potentials for crevice corrosion and stress corrosion cracking as functions of pH, ionic concentration, temperature, and metallurgical states from fabrication processes. Examples are presented to show how the approach can be used in system-level assessment of repository performance.     

美国放射性废物最终处置政策及技术路线

反应堆安全和核废物安全处置被认为是影响今后核能事业发展的两大障碍。自１９８０年以来，美国颁布了３个关于核废物处置的政策法令，对放射性废物的安全处置的要求及责任做出了明确规定。文章介绍了美国放射性废物处置的政策、技术路线及现状。对乏燃料及高放废物的处置，美国采取了比较慎重的态度，进行了各种方案的比较，虽然已有基本轮廓，但仍在探索之中。文章还介绍了高放废物处置中存在的一些有争议的重大问题和倾向性意见。     

Treatment and geological disposal of waste from NET pre-design

Within the European Fusion Technology programs Studsvik RadWaste AB has performed studies on fusion waste treatment and disposal for several years. This paper deals with the treatment and geological disposal of radioactive waste from NET operation and decommissioning. Results from calculations on radioactive waste fluxes for the operation and decommissioning of NET are reported. The calculations are based on the NET predesign report published 1993 and include results for the exchangeable in-vessel and external parts of the machine as well as permanent reactor components. Different aspects of treatment, packaging, transportation, and interim storage of the waste are discussed. The volumes of waste conditioned for final disposal are preliminarily quantified, according to German and Swedish scenarios for radioactive waste disposal. A total repository volume of approximately 45,000 m³ is required in the German Scenario and 35,000 m³ is required in the Swedish Scenario. Results from dose rate calculations for NET waste in final repositories are presented for the Swedish Scenario. This work was financially supported by the Swedish Natural Science Research Council (NFR) and the European Atomic Energy Community, under an association contract between Euratom and Sweden.     

高放废物地质处置库预选场址水文地质研究进展

深地质处置是目前国际上普遍接受的高放废物最终处置方案。对于这种处置方案而言，最有可能使处置库系统中放射性核素释放并进入生物圈的机制是地下水的作用。本文阐述了这种地下水的作用，包括地下水与工程屏障的相互作用、地下水在地质屏障中的核素迁移作用及核素滞留作用；介绍了处置库场址评价中水文地质研究的国际进展和动向；重点介绍了我国高放废物处置库预选场址水文地质研究进程和概况。     

Yucca Mountain engineered barrier system corrosion model (EBSCOM)

A revised engineered barrier system model has been developed by the Electric Power Research Institute to predict the time dependence of the failure of the drip shields and waste packages in the proposed Yucca Mountain repository. The revised model is based on new information on various corrosion processes developed by the US Department of Energy and others and for a 20-mm-thick waste package design with a double closure lid system. As with earlier versions of the corrosion model, the new EBSCOM code produces a best-estimate of the failure times of the various barriers. The model predicts that only 15% of waste packages will fail within a period of 1 million years. The times for the first corrosion failures are 40,000 years, 336,000 years, and 375,000 years for the drip shield, waste package, and combination of drip shield and the associated waste package, respectively.     

Design and thermal modeling of radially configured nuclear waste packages

A series of radial design configurations for packaging nuclear wastes are described. These radial arrangements for used nuclear fuel assemblies in containers are effective techniques for packaging significantly more radioactive waste in the available internal container volume. The radial package designs can be applied to packaging the nuclear waste for permanent storage at the Yucca Mountain (YM) repository. The radially configured containers will have high degree of structural strength and will be efficient in transferring heat from the waste form to the package surface due to the minimization of internal gaps. Radial configurations are reported for packaging the Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) used fuel assemblies. These configurations can be varied for co-packaging the colder, i.e. vitrified high level waste (HLW), canisters. Details of the geometry and the materials selected are discussed. Thermal analysis of the radial designs was conducted which confirm the feasibility of the designs demonstrating that no over-heating occurs in the contained nuclear waste in spite of the significantly extra amount of waste. The larger amount of packaged waste per container coupled with efficient heat transfer characteristics of these designs favor hotter and drier conditions for container surfaces in the YM emplacement drifts.     

The economic effects of the deferred disposal of spent fuel in Korea

The present study analyzes the economic effects concerning deferred disposal of spent fuel through long-term storage. According to the cost analysis, a scenario that a 90-year deferral of an HLW (High-Level Waste) repository construction in favor of a long-term storage of spent fuel would be economically preferable to another scenario based on the year 2040 chosen as the starting point for construction on a repository. That is, the former scenario would cost about 1/2 of the latter. This finding is an estimated result from an economic perspective only, assuming the disposal of 20,000-ton PWR spent fuel and 16,000-ton CANDU spent fuel. Still, it seems necessary to elicit proper term of storage for radioactive waste in order to comply with the so-called Polluter-Pays principle that the current generation cannot pass on its radioactive waste to the next generation.     

美国先进核燃料循环计划概述及对我国的启发

美国已重启核能计划,其中先进的核燃料循环计划是核能计划的核心。若这一计划得以顺利实施,将可以消除人们曾担忧的核能开发中的三大问题,核扩散、高放废物的处置和铀资源的可持续性问题。美国这一计划对我国的核能发展应有一些启发。     

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.     

Chapter 3: Type and Quantity of Radioactive Waste Destined for Disposal

Abstract

According to the current planning status, radioactive waste with negligible heat generation is destined for storage in the Konrad waste repository in accordance with the Provisional Waste Acceptance Requirements⁽⁴⁾. Waste of this kind occurs in the nuclear fuel cycle, research, medicine and technology. In its original state, the primary wastes of this type have various forms, such as: liquids, concentrates, sludges; ion exchange resins; compressible and/or combustible solids; incompressible solids, e.g. structural material components; filters, filter candles; other types of waste. The radioactive waste is appropriately conditioned and packaged before transport to the waste disposal site. To meet the basic requirements specified in the Waste Acceptance Requirements, primary waste must be solidified, for which the most common solidifying agents are cement and concrete although bitumen is also used. Waste products in decomposing, fermenting or liquid form or which contain a significant fraction in such states are not accepted for disposal.     

Progress towards a Deep Repository for Spent Nuclear Fuel in Sweden

The Swedish Nuclear Fuel and Waste Management Co. has in operation a safe and well integrated system for handling of all radioactive residues within Sweden. The existing central repository for low- and medium-level waste (SFR) and the central interim-storage facility for spent nuclear fuel (CLAB) can accommodate all the radioactive waste produced inside Sweden. Comprehensive research, development and demonstration activities are well under way for an encapsulation plant and a deep repository for spent fuel. These two facilities remain to be constructed to complete the waste management system. Siting of the deep repository is in progress with the aim of finding a suitable and accepted site. Implementation of the deep geological repository is a technical, scientific, social and political challenge. Smooth implementation must take into consideration both facts and emotions. Patience, flexibility and respect for the democratic process are important keywords. Research facilities, such as the underground Äspö Hard Rock Laboratory and the Encapsulation Laboratory, are important to promote scientific understanding as well as to demonstrate the disposal concept and technology.     

The kinetics of alpha-decay-induced amorphization in zircon and apatite containing weapons-grade plutonium or other actinides

Zircon and apatite form as actinide host phases in several high-level waste forms and have been proposed as host phases for the disposition of excess weapons-grade Pu and other actinides. Additionally, closely-related structure types appear as actinide-bearing phases among the corrosion products of spent nuclear fuel and high-level waste glasses. Self-radiation damage from -decay of the incorporated Pu or other actinides can affect the durability and performance of these actinide-bearing phases. For both zircon and apatite, these effects can be modeled as functions of storage time and repository temperature and validated by comparison with data from natural occurrences. Natural zircons and apatites, with ages up to 4 billion years, provide abundant evidence for their long-term durability because of their wide spread use in geochronology and fission-track dating. Detailed studies of natural zircons and apatites, ²³⁸Pu-containing zircon, a ²⁴⁴Cm-containing silicate apatite, and ion-irradiated zircon, natural apatite and synthetic silicate apatites provide a unique basis for the analysis of -decay effects over broad time scales. Models for -decay effects in zircon and apatite are developed that show -decay of Pu and other actinides will lead to a crystalline-to-amorphous transformation in zircon, but not in apatite, under conditions typical of a repository, such as the Yucca Mountain site.     

Concept of a Korean Reference Disposal System for Spent Fuels

A deep geologic disposal system for the spent fuels from nuclear power plants has been developed since this program was launched in 1997 in Korea. In this paper, the concept of a Korean reference high-level waste (HLW) vertical disposal system (KRS-V1) is described. Though no site for the underground repository has yet been specified in Korea, a generic site with a granitic rock is considered for a reference spent fuel repository design. The depth of the repository is assumed to be 500 m. The repository consists of a disposal area, a controlled area, and an uncontrolled area. The disposal area consists of disposal tunnels, panel tunnels, and a central tunnel. In the controlled area and the uncontrolled area, there are technical rooms and tunnels and/or shafts to connect them to the ground level, respectively. The repository will be excavated, operated, and backfilled in several phases including an underground research laboratory (URL) phase. The result of this concept development will be used for an evaluation of its feasibility, analyses of its long-term safety, information for public communication, and a cost estimation, among others.     

Preliminary Work Related to the Transport of Low and Intermediate Level Radioactive Waste in the Republic of Croatia

Abstract

In the Republic of Croatia, there is a project for a repository for low level (LL) and intermediate level (IL) radioactive waste disposal. Among many preliminary proceedings related to the proposed construction of the repository (such as site selection, repository project design, public acceptance, and the like), are the problems related to the transport of LL/IL radioactive waste from the place of its generaiion or storage to the location of the final disposal site. In this phase of the preliminary works— prior to site selection and further working out of project documentation for the facility, it has only been possible to commence with some study papers related to the problem. During 1992/1993 the first version of the generic study related to transport was worked out with the aim of preparing for a more detailed study, and technical and investment support for further working out of the project. The study is firstly a literature abstrac,, with an analysis of the transport problem for Croatia and recommendations for a transport system, type of transport, transport equipment and dynamics for the requirements of the Repubiic of Croatia.     

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.