Behavior of HPC nuclear waste disposal structures in leaching environment

Nuclear waste disposal is a subject of concern for the French national agency for nuclear waste management (Andra). One of the solutions envisaged by Andra (in accordance with the French law (June 28th 2006)) to take care of nuclear wastes is long-term underground storage. The storage structures would be high-performance concrete tunnels. The safety level of these structures for the next 1000 years has to be demonstrated by Andra. To do this, scenarios compatible with the geological and hydrological environment of the tunnel will be considered including chemical and physical evolutions. Among them is water saturation of the disposal structure leading to a concrete leaching process combined with a ground convergence towards the concrete vault, which is studied here. The corresponding research program is presented; it includes chemical and chemo-mechanical aspects and a simulation of the long-term behavior of disposal cell.     

Radiation damage of alkali borate glasses for application in safe nuclear waste disposal

The Electron Paramagnetic Resonance technique has been used to study the time decay of paramagnetic species induced by gamma irradiation and the radiation hardness of different alkali borate glasses for their application in safe nuclear waste disposal. Glasses with different composition have been prepared by conventional melt-quenching. Glass compositions have been chosen to elucidate the role of different alkali cations and of aluminium oxide on the borate glass network. The paramagnetic states detected in these glasses have been attributed, according to the literature, to the formation of hole centers associated with threefold coordinated boron. The results indicate that the time decay trend of the different glasses is slow and that the constant decay does not appear related to the chemical composition. Moreover, the undesired strong fading of the radiation-induced signal during the first 24 h after irradiation, observable in silicate glasses has not been detected. Although no species detectable by a X band spectrometer have been generated, the interaction of lithium borate glasses with air seem to accelerate the system decay rate. Annealing was finally performed and optimized, investigating the correlation between the chemical composition and the radiation damage recovery.     

The Canadian approach to safe, permanent disposal of nuclear fuel waste

The assessment of nuclear fuel waste disposal deep in plutonic rock of the Canadian Precambrian Shield is now well advanced. A comprehensive understanding has been developed of the chemical and physical processes controlling the containment of radionuclides in used fuel. The following conclusions have been reached:

• - Containers with outer shells of titanium or copper can be expected to isolate used fuel from contact with groundwater for at least 500 years, the period during which the hazard is greatest.

• - Uranium oxide fuel can be expected to dissolve at a rate less than 10⁻⁸ per day, resulting in very low rates of radionuclide release. This is consistent with observations of uranium oxide deposits in the earth's crust.

• - Transport of radionuclides away from the containers can be significantly delayed by placing a compacted bentonite-clay based layer between the container and the rock mass.

• - The granite plutons of interest consist of relatively large rock volumes of low permeability separated by relatively thin fracture zones. The low permeability volumes are sufficiently large to accommodate a vault design that will ensure radionuclides do not reach the surface in unacceptable concentrations.

Our field and laboratory investigations, together with assessments of conceptual disposal vault designs, give us confidence that the combination of engineered barriers and a technically suitable plutonic rock site will meet the requirements for safe disposal of nuclear fuel wastes in Canada.     

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.     

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.     

The Belgium underground research facility: Status on the demonstration issues for radioactive waste disposal in clay

Safe disposal of radioactive waste is one of the key issues in the consolidation process of the nuclear industry. Disposal in deep geological formations is at present the most promising option for long-lived waste and spent fuel. In Belgium, from a survey of potential geological formations, preference was expressed for a deep clay formation (Boom clay) lying under the facilities of the Mol Research Centre. This does not anticipate at all the site selection procedure. Rather early in the programme (1980) an underground laboratory was constructed in the clay to evaluate feasibility aspects and to become an in situ facility for performing tests in close to real conditions within the HADES (High Activity Disposal Experimental Site) project. The present programme requires now a major extension of the present underground facility to perform new large-scale in situ tests. A major action to be developed for the next 10 years is the PRACLAY project, a demonstration experiment simulating the thermal output of a 30 m long high level waste disposal gallery, 2 m in diameter. The experiment will be installed from an extension of the existing facilities to be built over the next 3 years. The experiment is planned to last until 2005.     

放射性废物处置安全若干进展

结合2005年放射性废物处置安全国际大会反映出来的放射性废物处置安全领域的最新进展,介绍了全球放射性废物安全框架、废物处置安全战略、安全方案、地质处置设施安全、近地表处置设施安全、中等深度废物处置方案和公众沟通等方面的若干新进展和新观点。     

Physico-chemical characterization of clays at Kalpakkam nuclear plant site towards assessment of radioactive waste disposal

In order to evaluate the safety of a radioactive waste repository in a geological formation, it is necessary to study the prevailing sub surface geology surrounding the disposal facility. The Kalpakkam subsurface was found to comprise of charnockite/hypersthene granite which is overlain by clay, sandy clay and sand. Clays can act as a barrier for groundwater flow and attenuate radionuclide migration. Accordingly, in the present study, clay samples were collected at chosen depths in the subsurface within the study area at Kalpakkam and colloidal clay fractions were separated. The clay colloids after separation were characterized for their size, shape, morphology, elemental composition and zeta potential. Photon correlation spectroscopy, scanning electron microscopy, energy dispersive X-ray analysis, X-ray diffraction technique and zetasizer were used to characterize the colloidal particles. The detailed characterization and analysis has revealed that the major clay mineral present in the subsurface at Kalpakkam is kaolinite. Minor quantities of smectite and illite were also identified in some clay samples. It was observed that the clays present in the study area are having low cation exchange capacity for radionuclides and emphasizes the requirement of proper backfill materials around disposal facilities for retardation of radionuclide migration. The occurrence of clay minerals at depths within the aquifer thickness warrants analysis of groundwater samples for suspended colloidal clay content. These suspended clay particles in groundwater have potential to transport of radionuclides in the ground environment along with groundwater flow depending on the nature of clay colloids and its zeta potential. Above investigations with respect to clay characterization would help the designer in incorporating the adequate safety measures in construction of near-surface disposal facility.     

地球化学工程学在放射性废物处置中的应用

介绍了应用地球化学工程学治理环境的基本依据,常用的放射性废物处置工程模式和工程屏障的功能,并以某放射性废物处置场地球化学工程屏障物料研究为例,说明地球化学工程学在放射性废物处置中的应用。研究结果表明,采用地球化学工程学方法来改良放射性废物处置场址的天然缺陷,可大大提高放射性废物处置的安全性。     

The inhibiting effects of hydrogen on the corrosion of uranium dioxide under nuclear waste disposal conditions

A number of electrochemical experiments were employed to investigate the effects of hydrogen on the corrosion of UO₂ under nuclear waste disposal conditions. A combination of corrosion potential (E_CORR) measurements and cyclic voltammetry have indicated that dissolved hydrogen can polarize the UO₂ surface to reducing potentials; i.e., to E_CORR values more negative then those observed under anoxic (argon-purged) conditions. A comparison of the behaviours of SIMFUEL specimens with and without incorporated noble metal ε-particles indicates that these particles may act as catalytic electrodes for H₂ oxidation, H₂ ↔ 2e⁻ + 2H⁺. It is the galvanic coupling of these particles to the UO₂ matrix which suppresses the fuel corrosion potential.     

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.     

高放废物地质处置场址安全要求

概述了国际原子能机构、美国、法国、瑞典、芬兰和日本的高放废物地质处置场址安全要求的内容、现状和发展趋势。场址安全要求主要从水文地质、地球化学、岩石特征、气候变化、人类干扰等方面来阐述场址的有利条件、不利条件以及潜在的有利与不利条件。以系统科学为指导,初步探讨了我国场址安全要求的研究框架和实施步骤。     

高放废物地质处置性能评价

为建立我国高放废物地质处置性能评价方法而系统地介绍了性能评价的研究目的、研究内容、研究方法、国内外研究现状；以此为基础，提出了关于开展我国性能评价的若干建议。性能评价方法的建立将有利于我国高放废物地质处置事业的协调发展。     

核废料核素价值研究

迄止21世纪初，全世界30多个国家和地区的450多座核电站在运行，为全球提供的电力超过总电力的16%。我国核电起步较晚，但到2010年，投入运行的核电机组也将超过20个。全世界已运行的核电机组绝大多数为轻水堆。因此，轻水堆电站，特别是压水堆(PWR)核电站发展中提出的问题将在很大程度上左右裂变核能的持续发展。经过多年的研究和发展，商业规模的压水堆核电站已可安全可靠、经济高效地运行。     

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.     

On risk assessment of high level radioactive waste disposal

Since the first controlled chain reaction, the use of nuclear fission to produce power has grown rapidly. One of the major concerns with the continued growth of the nuclear power industry is the production of the high level radioactive wastes which are by-products of the fission process. The risks associated with the disposal of high level wastes derive from the potential for release of radioactive materials into the environment. The assessment of these risks requires a methodology for risk analysis, an identification of the radioactive sources, and a method by which to express the relative hazard of the various radionuclides that comprise the high level waste.The development of a methodology for risk analysis is carried out after a review of previous work in the area of probabilistic risk assessment. The methodology suggested involves the probabilistic analysis of a general accident consequence distribution. In this analysis, the frequency aspect of the distribution is treated separately from the normalized probability function. In the final stage of the analysis, the frequency and probability characteristics of the distribution are recombined to provide an estimate of the risk.The characterization of the radioactive source term is accomplished using the ORIGEN computer code. Calculations are carried out for various reactor types and fuel cycles, and the overall waste hazard for a projected 35 year nuclear power program is determined. An index of relative nuclide hazard appropriate to problems involving the management of high level radioactive wastes is developed in this work. As an illustration of the methodology, risk analyses are made for two proposed methods for waste management: extraterrestrial disposal and interim surface storage. The results of these analyses indicate that, within the assumptions used, the risks of these management schemes are small compared with natural background radiation doses.     

桶装核废物的非破坏性分析(续)

有源γ射线(X射线)法可进一步分为吸收谱(吸收测量)法和有源诱发射线法。吸收谱法的分析原理是：当一定量的射线透射样品时,样品内的各元素具有正好吸收某一能量的射线的特性,从而使透射的射线强度大大减弱,这种减弱的多少与该元素的含量有关,这一现象称为     

桶装核废物的非破坏性分析

综述了桶装核废物非破坏性分析常规方法(分段γ扫描法、无源中子法、有源中子法、有源γ射线法)及基本原理、研究与进展的最新概况,介绍了生产相关仪器和设备的主要厂商。