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.     

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。     

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.     

Corrosion issues in nuclear waste disposal

This paper summarized some corrosion issues specific to nuclear waste disposal and illustrates them by the French geological clay concept for the reliable prediction of container degradation rate and engineering barrier integrity over extended periods, up to several thousands years. Among the items, the following are included:

• The importance of the underground repository conditions.

• The necessity of developing comprehensive semi-empirical models and also predictive models that must be based on the mechanisms of corrosion phenomena.

• The use of archaeological artefacts to demonstrate the feasibility of long term storage and to provide a database for testing and validating the models.

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.     

Nuclear waste management in the US: The Nuclear Waste Technical Review Board's perspective

The US program for the management and disposal of commercial spent nuclear fuel and high level waste is in a period of potential programmatic, regulatory, and legislative change. Proposals currently being considered by the US Congress would authorize the development of a storage facility as soon as possible adjacent to the potential repository site at Yucca Mountain. The legislation also would establish regulatory requirements for a permanent repository at an individual dose limit of 1 mSv year⁻¹ (100 mrem year⁻¹) for the average person living near the repository. Concurrently, the fiscal year 1996 appropriation to characterize the Yucca Mountain site has been reduced by approximately 40%. These initiatives portend possible changes in the focus of the US program, including a fundamental shift in priority from permanent disposal to temporary storage, and a change in the approach to licensing a potential repository at the Yucca Mountain site. This paper provides the perspective of the members of the Nuclear Waste Technical Review Board on the impact these developments could have on the future of the US program. It discusses the Board's opinion on how to address the issues these and other developments raise in a way which moves the US civilian radioactive waste management program forward.     

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.     

Development and validation of a CFD model predicting the backfill process of a nuclear waste gallery

Nuclear waste material may be stored in underground tunnels for long term storage. The example treated in this article is based on the current Belgian disposal concept for High-Level Waste (HLW), in which the nuclear waste material is packed in concrete shielded packages, called Supercontainers, which are inserted into these tunnels. After placement of the packages in the underground tunnels, the remaining voids between the packages and the tunnel lining is filled-up with a cement-based material called grout in order to encase the stored containers into the underground spacing. This encasement of the stored containers inside the tunnels is known as the backfill process.A good backfill process is necessary to stabilize the waste gallery against ground settlements. A numerical model to simulate the backfill process can help to improve and optimize the process by ensuring a homogeneous filling with no air voids and also optimization of the injection positions to achieve a homogeneous filling. The objective of the present work is to develop such a numerical code that can predict the backfill process well and validate the model against the available experiments and analytical solutions.In the present work the rheology of Grout is modelled as a Bingham fluid which is implemented in OpenFOAM - a finite volume-based open source computational fluid dynamics (CFD) tool box. Volume of fluid method (VOF) is used to track the interface between grout and air. The CFD model is validated and tested in three steps. First, the numerical implementation of the Bingham model is verified against an analytical solution for a channel flow. Second, the capability of the model for the prediction of the flow of grout is tested by means of a comparison of the simulations with experimental results from two standard flowability tests for concrete: the V-funnel flow time and slump flow tests. As a third step, the CFD model is compared with experiments in a transparent Plexiglas experimental test setup performed at Delft University of Technology, to test the model under more practical and realistic conditions. This experimental setup is a 1:12.5 scaled version of the setup of the full-scale mock-up test for backfilling of a waste gallery with emplaced canisters used in the European 6th framework project ESDRED (Bock et al., 2008). Furthermore, the plexiglas setup is used to study the influence of different backfill parameters.The CFD results for a channel flow shows good comparison against the analytical solution, demonstrating the correct implementation of the Bingham model in OpenFOAM. Also, the CFD results for the flowability tests show very good comparison with the experimental results, thereby ensuring a good prediction of the flow of grout. The simulations of the backfill process show good qualitative comparison with the plexiglas experiment. However, occurrence of segregation and also varying rheological properties of the grout in the plexiglas experiment results in significant differences between the simulation and the experiment.     

Radioactive waste management practices in India

Concern for the environment and establishment of radiation protection goals have been among the major priorities in planning of India's nuclear energy programme. In the Indian nuclear fuel cycle, right from inception, a closed loop option has been adopted where spent fuel is reprocessed to recover plutonium and unused uranium. The emphasis has been to recover actinides, individual fission products and recycle them back to the fuel cycle or use them for various industrial applications. The development of innovative treatment processes for low and intermediate level wastes in recent times has focused on volume reduction as one of the main objectives. In the case of high-level liquid waste, vitrification in borosilicate matrix is being practiced using induction heated metallic melters at industrial scale plants at Tarapur and Trombay.Currently, there are seven operating near surface disposal facilities co-located with power/research reactors in various parts of the country for disposal of low and intermediate level solid wastes. These are routinely subjected to monitoring and safety/performance assessment. An interim storage facility is operational for the storage of vitrified high-level waste overpacks for 30 years or more. Nation wide screening of potential regions and evaluation of rock mass characteristics is in progress for ongoing geological repository programme. Preliminary design and layout of an underground research laboratory/repository has also been initiated.A research programme is underway for long-term evaluation of vitrified waste product under simulated repository conditions. Research is also directed towards development of advanced technologies for waste processing as well as conditioning in vitreous and ceramic matrices. The Department of Atomic Energy with participation of the Indian industry has developed all essential remote-handling gadgets required for operation and maintenance of waste management system and assemblies including decommissioning.     

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.     

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.     

Mechanical properties of nuclear waste glasses

The mechanical properties of nuclear waste glasses are important as they will determine the degree of cracking that may occur either on cooling or following a handling accident. Recent interest in the vitrification of intermediate level radioactive waste (ILW) as well as high level radioactive waste (HLW) has led to the development of new waste glass compositions that have not previously been characterised. Therefore the mechanical properties, including Young’s modulus, Poisson’s ratio, hardness, indentation fracture toughness and brittleness of a series of glasses designed to safely incorporate wet ILW have been investigated. The results are presented and compared with the equivalent properties of an inactive simulant of the current UK HLW glass and other nuclear waste glasses from the literature. The higher density glasses tend to have slightly lower hardness and indentation fracture toughness values and slightly higher brittleness values, however, it is shown that the variations in mechanical properties between these different glasses are limited, are well within the range of published values for nuclear waste glasses, and that the surveyed data for all radioactive waste glasses fall within relatively narrow range.     

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。     

高放废物处置库围岩开挖损伤区研究的关键问题及进展

高放废物(HLW)地质处置是将高水平放射性废物埋存于地下500～1 000 m地质体中,使放射性废物与生物圈长期隔离。地质处置库对核素的长期隔离能力是安全评价的关键课题。地下硐室的开挖将不可避免地对围岩造成损伤,形成开挖损伤区(EDZ),改变围岩的物理力学特性,对高放废物地质处置长期安全性存在潜在的影响。目前多个国家建成了高放废物处置地下实验室,并开展了大型原位开挖损伤区的研究,研究开挖损伤区的形成过程及其物理力学特性的变化。本文综述了国外结晶岩地下实验室开展的开挖损伤区研究,总结了EDZ关键研究问题;梳理了加拿大、瑞典、芬兰3个地下实验室多年来开展的系统的EDZ研究工作,对当前EDZ预测模型及模拟技术进行了总结;对我国地下实验室将开展的开挖损伤区研究工作进行了初步探讨,期望为我国的相关研究提供借鉴。同时,高放废物处置库是地下工程新实践,其EDZ的研究成果,形成的技术方法将对其他行业地下工程的建设,如引水隧洞、公路铁路隧道等也有重要的参考价值。     

Status of the FRG activities on direct disposal of spent LWR-fuel

As stipulated by the German Atomic Energy Act, reprocessing is the reference waste management route for LWR's in the Federal Republic of Germany (FRG).Spent fuel disposal without reprocessing is being developed to technical maturity for those fuel elements for which reprocessing is either technically not feasible or economically not justifiable. The reference concept for direct disposal is the emplacement of large and heavily-shielded casks in drifts of a repository mine located in a salt dome. Moreover, a back-up solution is being pursued which results in smaller canisters which are emplaced in boreholes.The mining authorities have pointed out that the feasibility of direct disposal is to be demonstrated before a license for industrial scale deployment could be granted. Demonstration tests are necessary in the following areas: shaft transport of large and heavily shielded casks, handling of the casks in the repository and thermal and rock mechanics investigations with respect to the drift emplacement concept.The results of the demonstrations tests as well as the results from layout and optimization studies for a common repository for both reprocessing waste and spent fuel will be available early enough to be incorporated into the licensing procedure for the FRG's first repository for heat-generating nuclear wastes. This means that direct disposal of spent fuel not suitable for reprocessing could be introduced in the future in addition to the reprocessing and recycling waste management concept.     

A novel and inexpensive wet chemical route for precursor synthesis of synroc monoliths

Synroc waste forms containing a simulated nuclear waste of fast reactor origin were synthesized employing a simple and inexpensive wet chemical method by completely avoiding the use of organic reagents. Peroxytitanic acid prepared by dissolving titanium metal powder in an alkaline hydrogen peroxide solution was used as source for titania; the other constituents were taken as metal nitrates. A nano precursor with high surface area was obtained by co-precipitation, and it gave a dense synroc monolith on sintering. The precursor and final product were characterized by TG–DTA coupled with EG-MS, XRD, SEM and BET.     

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.     

Impact of thermal constraints on the optimal design of high-level waste repositories in geologic media

An approximate, semi-analytical heat conduction model for predicting the time-dependent temperature distribution in the region of a high-level waste repository has been developed. The model provides the basis for a systematic, inexpensive examination of the impact of several independent thermal design constraints on key repository design parameters and for determining the optimal set of design parameters which satisfy these constraints. Illustrative calculations have been carried out for conceptual repository designs for spent pressurized water reactor (PWR) fuel and reprocessed PWR high-level waste in salt and granite media.     

Long-term lining performance - Civil engineering problem of potential retrieval of buried spent nuclear fuel

The current solution for the spent fuel, high-level and long-lived radioactive waste is to store them at surface facilities from which they will be subsequently moved to a deep repository. No such repositories are in operation currently but several such facilities are close to the construction phase. A deep repository can be situated in several types of geological conditions including clay formations, salt sediments, argillites and tuffitic and granitic rocks. The character of the host rock is the key factor determining the design and specific requirements of individual components of such a facility. The future potential retrieval of canisters containing nuclear waste from the repository is a further influential factor. The reason for retrieval of containers lies in the development of fast reactors and increased interest for spent fuel reprocessing. Naturally, the decision as to whether retrievability is technically feasible must be made before finalising the design and construction process of the repository. If the decision is made to retrieve, a design which will include all the relevant safety aspects for the potential retrieval of canisters must be determined. The lay-out of the repository, the materials to be used and the design of the various structures of the facility (e.g. access tunnels, disposal shafts, buffer and backfill) are not the only issues to be addressed. The long-term stability of the system as a whole, i.e. of all the components, is crucial. Depending on the disposal concept chosen, the thermal load generated by the waste in the disposal container, saturation by water from the surrounding environment and the loading of the host rock massif will constitute the main processes which will affect the behaviour, safety and future functioning of the repository from the civil engineering point of view. The long-term stability of the lining of disposal galleries is a basic precondition for the safe removal of spent nuclear waste from deep underground repositories. The stability problems of tunnel linings exposed to long-term thermal load have not yet been properly addressed and form the subject of the European TIMODAZ project (Thermal Impact on the Damaged Zone around a Radioactive Waste Disposal in Clay Host Rocks) and also supported by the “Complex System of Methods for Directed Design and Assessment of Functional Properties of Building Materials” project. This paper describes the design, construction and currently available results of a 1:1 scale “in situ” disposal tunnel model which has been built at the Josef Underground Educational Facility in the Czech Republic.     

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.