A simplified methodology for nuclear waste repository thermal analysis

A simplified model for repository thermal analysis is presented in this paper. The proposed model is to provide a general capability to efficiently calculate the time dependent temperature field in a geologic repository. The model analyzes both horizontal and vertical emplacement of nuclear waste packages. Verification of the code was performed based on the comparison with detailed numerical method-based standard models. The new model’s utility was demonstrated through a case study where a large number of repository-scale thermal analysis calculations is needed.     

Forced cooling of a nuclear waste repository mine drift—A scoping analysis

Nuclear waste repositories, with decay heat generation beneath the mine drift floors, are force-cooled with air so that re-entrance is possible many years after the waste has been buried. A numerical model has been developed which uses heat transfer coefficients as input. It has been demonstrated that mixed (forced and free) convective and surface roughness effects are significant and must be included in future experiments if reliable predictions are to be made of the time required to cool the repository. For example, when repository mine drifts in volcanic tuff are force-cooled, with forced convection being the only energy transport mechanism, it takes ≈ 0.1 year to cool the mine surface to a safe temperature. However, when mixed convection is the primary transport mechanism it takes ≈ 1.0 year to cool the mine. In addition to mixed convection, other effects are delineated.     

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.     

Requirements on waste packages resulting from mechanical and thermal impact during the operational phase of a waste repository

The aim of the disposal of radioactive waste is the protection of man and environment during the operational and the post-operational phase from the ionizing radiation of the radionuclides contained in the waste. In order to ensure this protection goal requirements are derived for the design of the repository as well as for the radioactive waste to be disposed of.The technical requirements for the design of the plant include for example the closing-off structure of filled emplacement rooms. The requirements for waste packages derived from safety analysis are related to the activity inventory, the waste form and the packaging. The type and extent of these requirements are described and it is shown on which safety aspects they are basedAs an example the radioactive waste with negligible heat generation is considered. The requirements on six different waste forms and on two waste classes of the packaging result from the mechanical and thermal impact during normal operation and in the case of incidents and serve as a base for the conditioning of the radioactive waste. Moreover, the way in which the properties of the waste form and the packaging are taken into account in the long-term safety analysis is described.     

Radioactive releases from nuclear waste repository in the tropical climate environment

The spent nuclear fuel from any nuclear power plant poses danger to the public if not taken care sufficiently. This report discusses the potential releases of radioactive materials to the accessible environment when the waste form is in the format using Synroc-C developed by ANSTO. The environment under the study is tropical environment and the repository area is below the groundwater table. The computational model was developed and run using the computer code called Repository Integration Program (RIP) developed by the Golder Associates. The result is considered a very low release to the accessible environment     

Numerical aspects of the study of the regional thermal impact of a radioactive waste repository

Numerical results are obtained for the regional thermal history of a radioactive waste repository. For such a purpose, an alternating direction implicit method has been used, with special treatment of singularities in heat generation, with a variable grid calculation. The results obtained are in accordance with previous results in calibration cases. These results show that the technique developed constitutes a valuable tool for the handling of problems such as the one under consideration. Its application is exemplified by means of a case with a typical geometrical configuration and thermal load.     

高放废物地质处置库热学分析的参数敏感性研究

应用FLAC3D软件建立高放废物地质处置库热学分析的简化计算模型,选择影响处置库温度场的包括材料热学参数、几何参数以及时间参数在内的16个关键参数,以膨润土内表面峰值温度(该物理量是高放废物地质处置库热学设计计算中作为温度准则的物理量)为参数敏感性分析的目标物理量,通过热学计算开展参数敏感性分析。在参数敏感性分析中,将参数敏感程度划分为高、中、低三等。分析表明:4个参数(膨润土导热系数、膨润土厚度、围岩导热系数、高放废物中间贮存时间)为高敏感度参数,2个参数(散热材料厚度、回填材料厚度)为中度敏感性参数,其它10个参数(高放玻璃固化废物体、外包装容器、散热材料、回填材料的导热系数与比热,以及膨润土与围岩的比热)为低敏感度参数。通过分析可以得到如下结论:在设计高放废物地质处置库时,对膨润土及围岩导热系数的测试应力求准确,对测试结果数据认真分析,确保为设计计算提供合理的输入参数;在确保膨润土满足工艺要求功能的前提下,宜尽量减小膨润土的厚度;按照本文热学分析模型初步估算,我国高放废物至少需要中间贮存20 a以上。     

The precision of product consistency tests conducted with a glass-bonded ceramic waste form

The product consistency test (PCT) that is used for qualification of borosilicate high-level radioactive waste (HLW) glasses for disposal can be used for the same purpose in the qualification of the glass-bonded sodalite ceramic waste form (CWF). The CWF was developed to immobilize radioactive salt wastes generated during the electrometallurgical treatment of spent sodium-bonded nuclear fuels. An interlaboratory study was conducted to measure the precision of PCTs conducted with the CWF for comparison with the precision of PCTs conducted with HLW glasses. The six independent sets of triplicate PCT results generated in the study were used to calculate the intralaboratory and interlaboratory consistency based on the concentrations of Al, B, Na, and Si in the test solutions. The results indicate that PCTs can be conducted as precisely with the CWF as with HLW glasses. For example, the values of the reproducibility standard deviation for Al, B, Na, and Si were 1.36, 0.347, 3.40, and 2.97 mg/l for PCT with CWF. These values are within the range of values measured for borosilicate glasses, including reference HLW glasses.     

Transient thermal response in nuclear waste repositories

In this paper, it is shown that a previously reported non-linear, one-dimensional, theoretical approximation simplifies — from a computational point of view — the calculation of the time-decay temperature field in nuclear waste repositories (NWR). This conclusion has been reached after solving, by using the control volume numerical method, the full three dimensional, transient, non-linear heat diffusion equation. The transient thermal field in a rock salt repository, is analytically solved and numerically predicted, along 100 years, after the disposal of a high-level waste (HLW). The nuclear waste, with a half-life of 32.9 years, releases an exponentially time dependent heat flux with 12 W m⁻² as the initial thermal load. Two cases are studied, in the first one it is assumed that the conductivity (k) and the volumetric heat capacity ρc_p of the host rock (diffusion domain) remain constant (linear case), whereas in the second one, a more realistic situation is analysed. In this last case, the conductivity of the rock salt varies as a function of the temperature field and the product ρ×c_p remains constant (non-linear case). In order to observe the effect of the salt conductivity (constant or variable) on the repository temperature distribution, a comparison of both cases is performed. It is concluded, that the theoretical model, which provides an analytical solution of the thermal fields may be a powerful low cost method for design purposes.     

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.     

高放废物地质处置中的工程材料

凡人类从事于与核材料有关的许多生产、生活活动均可能产生不同活度的放射性废物.高放废物由于具有放射性水平高、发热量大、核素寿命长等特点,其安全处置倍受全球科学家和广大公众所重视.目前深地质处置被国际上公认为处置高放废物的最有效可行的方法.借鉴已有研究成果,我国采用多重工程屏障系统(包括废物固化体、废物罐及其外包装和缓冲/回填材料)和适宜的地质围岩地质体共同作用来确保高放废物与生物圈的安全隔离.参照国际上该领域的研究成果,结合我国处置概念,本文就高放废物地质处置中的工程材料(废物固化体、废物罐、外包装、缓冲材料、回填材料),以及其材料选择、设计要求和研究重点等进行了总结.     

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.     

The transport of radionuclides discharged from a geologic waste repository

A 3-D conservative solution to the convection-dispersion-retardation equation describing the migration of radionuclides from a geologic nuclear waste disposal site is derived using a Green's function approach under conditions in which the contaminant inventory is preferentially discharged to the steady groundwater flow. The numerical implementation of the upper estimate to the transported inventory, as embodied in the CANUKE© code, is described; its applicability is demonstrated by calculating the groundwater concentration of the neptunium series nuclides released from used CANDU™ fuel emplaced in an hypothetical repository; and its general usage is discussed. The ability to handle actinide chains of arbitrary length, whatever their initial inventories, yields both less conservative and physically more realistic estimates of transported inventories than schemes which begin with a contracted chain of longer-lived nuclides whose progeny are assumed to accompany them in their migration, irrespective of their relative mobilities.     

Sensitivity calculations for low-heat generating defense waste repository temperatures

Repository temperatures were calculated for Savannah River wastes by using both two- and three-dimensional numerical schemes. The error introduced by using the simpler and more efficient two-dimensional models is less than the present uncertainties introduced by waste power generation and host rock properties. Waste canister temperatures were found to be relatively insensitive to geometric asymmetry and model detail outside the immediate vicinity of the waste canister.     

Liquidating a hard-to-reach high-level waste repository at the Kurchatov Institute

Details of organizing the work to liquidate a hard-to-reach repository for high-level waste at a special site at the Institute are described. The bulk of the waste in the pit was encased in a high-strength concrete slab and, together with low- and medium-level waste, contained a large number of metal cans of high-level waste. Special arrangements for radiation protection set up around the pit and the techniques used to break up the concrete casing and extract the waste are described. Video cameras and a gamma visualizer were used to find high-level waste and fragments in the demolished concrete casing and to guide remotely controlled robotic equipment to them. Changes in the radiation environment in the work area were monitored operationally with a gamma locator; data from this detector were fed in real time to and analyzed and processed by a personal computer for use in carrying out the work. Translated from Atomnaya énergiya, Vol. 105, No. 3, pp. 164–169, September, 2008.     

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.     

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.     

Some aspects of nuclear and radiological safety of operating a spent nuclear fuel repository

A model of an irregular situation in a spent nuclear fuel repository with the introduction of excess reactivity into the system, consisting of containers with spent fuel assemblies and water, is examined. The neutron kinetics of a critical system is calculated taking account of the thermohydraulics of the system. The character of the flow of a short-time self-sustained chain reaction — “neutron burst” — is described. It is found that an excursion of the system in the range of reactivity introduction rates examined will result in heating of the system and self-quenching of the chain reaction by negative reactivity effects with respect to fuel temperature. Intense fluxes of fission neutrons and prompt gamma rays, accompanying a self-sustained chain reaction, are formed in the excursion process. A mixed neutron and gamma ray field near the system considered is investigated. __________ Translated from Atomnaya énergiya,Vol. 104, No. 3, pp. 141–147, March, 2008.     

高放废物地质处置库花岗岩体预选研究

简要介绍了应用遥感技术结合地质研究,优选高放废物地质处置库花岗岩体的思路和方法.以北山外围地区为例,通过遥感数据处理,遥感影像解译,岩体地质特征分析,岩体预选准则的建立,在野外勘查的基础上,预选了若干有利的花岗岩体.为优选高放废物地质处置库场址提供决策依据.     

Preliminary thermal hydraulic safety analysis of water-cooled ceramic breeder blanket for CFETR

The Chinese fusion engineering test reactor (CFETR) was expected to bridge from the international thermonuclear experimental reactor (ITER) to the demonstration fusion reactor (DEMO). The water-cooled ceramic breeder (WCCB) blanket is one of the blanket candidates for CFETR. In this paper, preliminary thermal hydraulic safety analyses have been carried out using the system safety analysis code RELAP5 originally developed for light water fission reactors. The pulse operation and three typical loss of coolant accidents (LOCAs), namely, in-vessel LOCA, in-box LOCA, and ex-vessel LOCA, were simulated based on steady-state initialization. Simulation results show that important thermal hydraulic parameters, such as pressure and temperature can meet the design criterion which preliminarily verifies the feasibility of the WCCB blanket from the safety point of view.