Disposal Of Spent Nuclear Fuel Research Articles

Overview of the Lithuanian programme for disposal of RBMK-1500 spent nuclear fuel

AbstractIn Lithuania all the spent nuclear fuel (SNF) came from operation of the Ignalina nuclear power plant with two reactors of RBMK type (RBMK is a Russian acronym for 'Channelized Large Power Reactor' which is a water-cooled graphite-moderated reactor: RBMK-1500). Approximately 22,000 SNF assemblies are due for geological disposal in Lithuania. Currently it is envisaged that SNF will be stored in dry interim storage facilities (new and existing) for at least 50 y prior to possible deep geological disposal.The decision on the final SNF management option (disposal in a national repository, disposal in regional repositories, etc.) has not yet been made but some investigations of the possibilities to dispose of the SNF in Lithuania have been initiated. With the support of Swedish experts, analysis of possible geological formations for SNF disposal was performed and the existence of potentially suitable formations agreed. The geological formations prioritized as prospective include the crystalline rocks in southern Lithuania and two clayey formations: the Lower Triassic clay formation and the Lower Cambrian Baltic Group clay formation, with priority given to the Lower Triassic clay formation.This paper presents the main aspects of the research and other activities undertaken over the past decade in the field of SNF disposal: international cooperation; current status and plans for the Lithuanian national program; further investigations required; and competence developments.

Retention of selenium and caesium on crystalline rock: the effect of redox conditions and mineralogical composition

Isotopes of caesium and selenium belong among fission products, being present in spent nuclear fuel (SNF) and high level waste (HLW), and contribute significantly to the risk associated with SNF and HLW disposal in deep geological repositories. The present work is focused on sorption behaviour of Cs and Se species on two types of crystalline rock, granite from Melechov massif (Czech Republic) and Aspo diorite from the underground laboratory Aspo (Sweden). The aim was to compare sorption behaviour of Cs and Se on these rocks under aerobic and anaerobic conditions and determine the influence of redox conditions on their kinetic sorption properties.

Ethics and the management of spent nuclear fuel

In March 2011, Swedish Nuclear Fuel Management Co (SKB) submitted an application to the Swedish Radiation Safety Authority (SSM), and the Nacka Land and Environmental court to build a repository for high-level spent nuclear fuel (SNF) in the municipality of Östhammar. The purpose of this paper is (1) to present the KBS-3 method for the direct disposal of high-level SNF and the debate about the method, (2) to analyze the ethical principles involved, and (3) how to resolve possible conflict between these principles. The paper is divided in three parts. Part 1 contains a presentation of the KBS-3 method and different items in the debate about it. Part 2 concerns the ethical dimension of the method and the different ethical principles that can be discerned explicitly or implicitly in the discussion and the regulatory framework that informs the discussion. Part 3 contains an analysis of the conflicting principles and how this conflict can be resolved. For example, it is argued that some kind of ethical meta-norm might be formulated and that the ethical principles involved in the management of SNF might be assessed in reference to such ethical meta-norm. Finally, it is discussed if and how the conflict between different principles for the management of SNF might be resolved through closer consideration of the relationship between ethical principles and technical practice.

Uncertainty and sensitivity analysis of radionuclide migration through the engineered barriers of deep geological repository: Case of RBMK-1500 SNF

Like other nuclear countries, Lithuania is facing the problem of spent nuclear fuel (SNF) disposal from RBMK11“RBMK” is a Russian acronym for “Channelized Large Power Reactor”.-1500 type reactors at the Ignalina NPP (~22,000 of the SNF assemblies). The application of probabilistic and deterministic methods for uncertainty assessment and important parameter ranking is of high importance in order to build confidence in the assessment results on radionuclide migration and indicates directions for further uncertainty reduction.This study is devoted to the uncertainty and sensitivity analysis of radionuclide migration through the engineered barriers of a conceptual geological repository for RBMK-1500 SNF disposal. The analysis was performed using computer codes AMBER (United Kingdom) and MATLAB (Sweden). The analysis showed that the uncertainty of defect enlargement time had no significant impact on 129I and 226Ra peak flux from the engineered barriers yet it was an important parameter for time dependent mean flux of 129I. Sensitivity analysis revealed that in the case of 129I peak flux, further investigations should be oriented to more precise determination of diffusion coefficient in bentonite, instant release fraction and SNF dissolution rate. In the case of 226Ra, efforts to reduce the uncertainty of SNF dissolution rate, sorption coefficient in bentonite and equivalent flow rate should be made.

Numerical modelling of radionuclide inventory for RBMK irradiated nuclear fuel

After the final shutdown of Ignalina NPP Unit 1 in 2004 and Unit 2 in 2009, total amount of spent nuclear fuel (SNF) is approximately 22,000 of fuel assemblies. Radionuclide content in SNF and its characteristics are initial data for analysis of various subjects such as radiation shielding, thermal analysis and other processes that can affect the performance of SNF storage or disposal facilities’ components. For instance, ageing of the components is influenced by decay heat, neutron and gamma radiation, external environment conditions, mechanical stresses, etc. The first two factors can be defined only after SNF characteristics are investigated and irradiation levels are determined. Experimental investigation of radionuclide content and characteristics in SNF is expensive and complicated, so numerical methods are widely used. In this paper, results of numerical modelling of spent RBMK-1500 nuclear fuel characteristics and its variation with time are presented. Modelling was performed using SCALE computer code system TRITON module. Some calculation results are compared with available experimental data for RBMK-1000 spent nuclear fuel and a rather good agreement between numerical and experimental results is identified. Comparison of calculated gamma and neutron sources, activities and decay heats shows that results for RBMK-1500 fuel with various enrichments differ by 10–20%.

Comparison of Radionuclide Releases from a Conceptual Geological Repository for RBMK-1500 and BWR Spent Nuclear Fuel

Approximately 22 600 spent nuclear fuel (SNF) assemblies originating from the RBMK-1500 reactor of the Ignalina nuclear power plant in Lithuania need to be managed and disposed of safely. Generic investigations of RBMK-1500 SNF disposal options in Lithuania were initiated. This paper presents insights on RBMK-1500 SNF disposal in crystalline rocks gained during participation in the International Atomic Energy Agency Coordinated Research Project “The Use of Numerical Models in Support of Site Characterization and Performance Assessment Studies for Geological Repositories,” as well as in the Lithuanian Science Development Program. The research was focused on the analysis of disposal behavior of different SNF types under generic geological conditions and for a one-canister defect scenario with two different corrosion rates. A comparison of peak fluxes from the near field for Lithuanian RBMK-1500 and Swedish boiling water reactor SNF revealed differences that are not directly proportional to the differences in SNF inventory.

Socio‐Technical Challenges of Finland's Nuclear Waste Policy: Discussion of the Techno‐Scientific Community on the Geological Disposal of Spent Nuclear Fuel

AbstractThe objective of the paper is to study different views on the future challenges that the geological disposal of spent nuclear fuel (SNF) will encounter. We investigate how engineers, natural scientists, representatives of the nuclear industry, regulators, and laypersons from non‐governmental organizations perceive the remaining socio‐technical challenges for implementing geological disposal by examining discussions in a professional magazine covering nuclear waste management issues. The paper focuses on the post‐site selection phase associated with the planned disposal facility at Olkiluoto in Finland. This phase (between 1999 and 2010) represents the period after the nuclear waste management company Posiva submitted the Decision‐in‐Principle (DiP) application for the final disposal facility for SNF in 1999 and the ratification of the DiP regarding the final disposal of SNF by the Finnish Parliament in May 2001. The paper provides a theoretical elaboration of the term socio‐technical that will lay the foundation for empirical analysis as well as a brief overview of the state of affairs regarding the Finnish SNF management program. Results show that market actors were clearly the most active players in defining and discussing final disposal. Qualitative analysis identified six kinds of socio‐technical challenges: socio‐technical regime challenges, social acceptance challenges, knowledge challenges, safety challenges, construction challenges, and flexibility challenges.

사용후핵연료 심지층처분장부지 지질환경 기본요건 검토

사용후핵연료 심지층처분장 부지선정과 최종 처분장부지의 처분적합성을 평가하는 업무는 시행-착오를 줄이고 기술적 신뢰성 확보와 합리적이고 효율적인 업무수행을 추구하여야 한다. 이에 선행하여, 우리나라에 적용 가능한 처분장부지의 지질환경 요건 설정을 위한 기본방향과 개별 인자의 처분적합성지표를 가능한 한 정량화하여 설정하고 업무에 적용하여야 한다. 사용후핵연료 처분장부지 선정과 최종처분장 부지의 안전성확보를 위한 처분요건과 관련하여 IAEA 및 OECD 회원국들과 처분연구 및 상용사업 수행 관련 선진국가들의 사례를 바탕으로 요건 별로 구분하여 현황을 분석하였다. 여기서는 사용후핵연료 처분장 부지로서 암석 암반이 갖추어야 할 충분 혹은 선호요건에 대한 이해 제고와 관련 세부 기술지침을 도출하는데 기여하고자 하였다. 이를 토대로 어떠한 암석 암반이 상대적으로 보다 유리한 조건을 가지는 선호요건으로 제시해야 하는지, 그리고 충분요건과 선호요건을 적용하여 후보부지 조사 선별평가 기간 동안 부지선정업무에 반영하고 평가하고 결정하여야 하는 방법론을 도출할 수 있도록 기본 골격을 제시하였다. 또한 처분안전성 확보를 위해 필요한 기본적인 사항을 검토하고 서술하였다. 본 논문에서 기술한 항목들은 처분안전성 확보를 위한 처분요건의 기술지침 구성 체계, 처분안전성 확보개념, 다중방벽 기능 조건, 천연방벽의 지질환경 기본요건, 그리고 우리나라에 적용 가능한 처분장부지 지질환경 기본요건(안) 등으로 구성된다. 우리나라의 사용후핵연료 심지층처분장 부지의 위치에 관한 사업자 기술지침 요건으로 제안하였다. 이와 관련하여 충분요건과 선호요건으로, 화산활동, 지진활동, 단층운동 융기 침강 운동 및 기후 해수면변동 등 장기지질안정성 요건을 비롯한 15개 충분요건과 48개 선호요건을 제안하였다. 이들 요건은 우리나라의 지질환경 특성을 충분히 반영하여 후속되는 각 부문별 특성에 적합한 정량적인 기술 기준 및 지침으로 개발되어야 할 것이다. 정량적 기술지침의 도출은 상용 처분장부지 선별평가과정 및 처분장 부지적합성평가 과정으로부터 확립될 수 있을 것이다. 또한 다양한 부문별 안전사례(safety case) 작성 혹은 연구용 지하처분연구시설 (underground research laboratory: URL)을 이용한 처분시스템의 실증과정 등을 통하여 객관적이고 신뢰성있는 정량적인 지침들이 확립될 수 있을 것이다. This paper gives some basic requirements and preferences of various geological environmental conditions for the final deep geological repository of spent nuclear fuel (SNF). This study also indicates how the requirements and preferences are to be considered prior to the selection of sites for a site investigation as well as the final disposal in Korea. The results of the study are based on the knowledge and experience from the IAEA and NEA/OECD as well as the advanced countries in SNF disposal project. This study discusses and suggests preliminary guideline of the disposal requirements including geological, mechanical, thermal, hydrogeological, chemical and transport properties of host rock with long term geological stabilities which influence the functions of a multi-barrier disposal system. To apply and determine whether requirements and preferences for a given parameter are satisfied at different stages during a site selection and suitability assessment of a final disposal site, the quantitative criteria in each area should be formulated with credibility through relevant research and development efforts for the deep geological environment during the site screening and selection processes as well as specific studies such as productions of safety cases and validation studies using a generic underground research laboratory (URL) in Korea.

고준위폐기물다발의 단면형상 변화에 따른 가압경수로(PWR)용 고준위폐기물 처분용기의 구조해석

A structural model of the SNF(spent nuclear fuel) disposal canister for the PWR(pressurized water reactor) for about 10,000 years long term deposition at a 500m deep granitic bedrock repository has been developed through various structural safety evaluations. The SNF disposal baskets of this canister model have the array type whose four square cross section baskets stand parallel to each other and symmetrically with respect to the center of the canister section. However, whether this developed structural model of the SNF disposal canister is optimal is not determinable yet. Especially, there is still a problem in weight-reduction of the canister. The cross section shape of the SNF basket should be changed to solve this problem. There are two ways in changing the cross section shape of the SNF basket; the one is to rotate the cross section itself and the other is to change the cross section shape as other shape different from the square cross section. The previous study shows that the canister with rotated basket array is structurally more stable than the canister with un-rotated parallel basket array. However, whether this canister with rotated basket array is optimal is not either determinable as yet, because it is not revealed that the canister with other cross section different from the square cross section is structurally more stable than other canisters. Therefore, the structural analysis of the SNF disposal canister with other cross section shape which is also symmetric with respect to the canister center planes is very necessary. The structural analysis of the canister with various cross section shape basket array in which each basket is arrayed symmetrically with respect to the center planes is carried out in this paper. The structural analysis result shows that the SNF disposal canister with circular cross section shape baskets located symmetrically with respect to the center of the canister section is structurally more stable than the previously developed SNF disposal canister with the parallel basket array.

Finite Element Analysis of Transient Heat Transfer in and around a Deep Geological Repository for a Spent Nuclear Fuel Disposal Canister and the Heat Generation of the Spent Nuclear Fuel

This paper presents a finite element analysis of transient heat transfer in and around a hypothetical deep geological repository for a spent nuclear fuel (SNF) disposal canister and the heat generation of the SNF inside the canister to provide basic information for dimensioning the repository and configuring the repository components. Three geometric models are compared to determine the most suitable assuming the periodic allocation of boreholes where canisters are deposited. These models consist of several different material regions. Each model is horizontally limited to a region around and including a single canister, bounded by midsurfaces with variant distances between adjacent deposition tunnels and adjacent canisters, and vertically bounded by the ground surface located 500 m above the deposition tunnel and the surface located 500 m below the bottom of the borehole. Using a commercial finite element analysis code and detailed realistic finite element models of repository components, transient heat transfer analyses are carried out for up to 1000 yr after deposition of the canister into the repository. Time-dependent temperature curves at selected positions are obtained for each geometric model. Various temperature distribution changes of material regions in geometric models are also obtained.

A Container Based on Polymer Composite Materials for the Ultimate Disposal of Spent Nuclear Fuel and Radioactive Waste

This work demonstrates the feasibility of fabricating containers for the ultimate disposal of spent nuclear reactor fuel and high-level radioactive waste using polymer-based composite materials. The study has identified three engineering polymers suitable for this demanding application: polyetheretherketone (PEEK), polyetherimide (PEI), and polysulfone (PSU). PEEK and PEI are used as composite materials components, with 30% carbon and glass fiber, respectively, whereas PSU is used as a virgin (nonreinforced) material. The rationale for the choice of polymer composites comes from their superior physical, mechanical, and chemical performance, in addition to their economical advantage. In particular, they display better resistance to corrosion and to structural weakening from irradiation.Scaled-down containers were fabricated using these materials. They were subjected to a battery of tests under conditions similar to those expected for the disposal environment of actual radioactive waste–filled containers. In particular, the container models were irradiated in the pool of a SLOWPOKE-2 nuclear research reactor, accumulating doses from a mixed-radiation field that were comparable to total doses accumulated over 500 yr at a deep underground waste repository site. Mechanical compression tests mimicked the large hydrostatic pressures incurred from granite rock at depths of some 1000 m within the Canadian Shield.Several composite materials were tested, and for the three engineering materials listed above, some of the results are as follows:1. variation in elastic modulus following a 28.9-kGy radiation dose—PEEK, −6.66% ± 0.47%; PEI, +5.63% ± 0.23%; PSU, +3.16% ± 0.13%2. compression results for the irradiated container models and load at break and strain—PEEK, 2.152 MPa and 1178 μmm-1; PEI, 1.236 MPa and 1171 μmm-1; PSU, 1.190 MPa and 2576 μmm-1, respectively3. cost analysis—costs for the fabrication of the prototype containers based on PEEK, $273610; PEI, $145920; PSU, $257460.The work also provided insight into potential problems in the fabrication of full-sized containers and into the best fabrication methods to adopt. The method of filament winding would be more appropriate for the PEEK- and the PEI-based composite materials, while blow forming would be the preferred method for the PSU material. In particular, this research could determine the best way to design the container lids.

Advanced Polymer Composites for the Fabrication of Spent Nuclear Fuel Disposal Containers

In Canada, the spent nuclear fuel disposal method proposed is to permanently isolate the spent fuel in deep underground vaults carved in stable granite rock formations within the Canadian Shield, with the integrity of the isolation to be assured for a minimum period of 500 yr. The present work aims at determining the feasibility of using a consolidated composite material made of an advanced polymer called PEEK (Poly Ether Ether Ketone) and continuous graphite fiber to fabricate a container designed to isolate the spent nuclear fuel from the biosphere for such very long time periods. The research focused on submitting the PEEK-based composite material to a thermal and radioactive environment comparable to, and, in some aspects, more aggressive than, the conditions of exposure in the disposal vault. The changes to the physical, mechanical, and chemical properties of the material following prolonged exposure were then determined. The simulation of the environment was achieved by irradiating numerous test specimens in a mixed radiation field produced by a SLOWPOKE-2 nuclear research reactor at controlled ambient temperatures ranging from ˜20 to 75°C. The specimens were characterized via several methods: tensile and flexural testing, differential scanning calorimetry, scanning electron microscopy, and wide-angle X-ray scattering. The results confirmed that the PEEK-based composite material was resistant to exposure to high radiation doses (1 MGy), at temperatures between ˜20 and 75°C. The mechanical and other properties were barely affected, with values rarely exceeding 1σ of the properties of nonirradiated samples, suggesting that the PEEK–graphite fiber composite material can indeed be considered as a very good candidate for this demanding application.

Investigation of Yucca Mountain repository capacity for the US spent nuclear fuel inventory

An analytical decay heat model was developed to evaluate the US inventory of commercial spent nuclear fuel (SNF). The model was benchmarked against the results from ORIGEN-ARP 5.01. The new analytical SNF decay heat model was applied to actual (thru 2002) and projected SNF data. The total decay heat from the 63,000 MT commercial SNF at year 2012 was estimated at 182 MW. According to the thermal loading analysis using a mountain-scale heat transfer model, a 4.9 km 2 (1165 acre) site designated for SNF disposal was found to have the capacity to store more SNF than the statutory limit of 70,000 MTIHM. The maximum capacity available for SNF disposal at the Yucca Mountain site is dependent upon the thermal loading strategy chosen and SNF cooling time before emplacement. It was also shown that using high burnup SNF and adjusting the drift spacing, the capacity of the repository could be maximized on a per energy production basis, although some additional cooling may be necessary. Future work needs to consider extending the ‘footprint’ of the repository, applying non-uniform SNF loading into the drifts, and the impact of spent fuel reprocessing and other decay heat management strategies.

Behavior of uranium under conditions of interaction of rocks and ores with subsurface water

The behavior of uranium during interaction of subsurface water with crystalline rocks and uranium ores is considered in connection with the problem of safe underground insulation of spent nuclear fuel (SNF). Since subsurface water interacts with crystalline rocks formed at a high temperature, the mineral composition of these rocks and uranium species therein are thermodynamically unstable. Therefore, reactions directed toward the establishment of equilibrium proceed in the water-rock system. At great depths that are characterized by hindered water exchange, where subsurface water acquires near-neutral and reducing properties, the interaction is extremely sluggish and is expressed in the formation of micro- and nanoparticles of secondary minerals. Under such conditions, the slow diffusion redistribution of uranium with enrichment in absorbed forms relative to all other uranium species is realized as well. The products of secondary alteration of Fe- and Ti-bearing minerals serve as the main sorbents of uranium. The rate of alteration of minerals and conversion of uranium species into absorbed forms is slow, and the results of these processes are insignificant, so that the rocks and uranium species therein may be regarded as unaltered. Under reducing conditions, subsurface water is always saturated with uranium. Whether water interacts with rock or uranium ore, the equilibrium uranium concentration in water is only ≤10−8 mol/l. Uraninite ore under such conditions always remains stable irrespective of its age. The stability conditions of uranium ore are quite suitable for safe insulation of SNF, which consists of 95% uraninite (UO2) and is a confinement matrix for all other radionuclides. The disposal of SNF in massifs of crystalline rocks at depths below 500 m, where reducing conditions are predominant, is a reliable guarantee of high SNF stability. Under oxidizing conditions of the upper hydrodynamic zone, the rate of interaction of rocks with subsurface water increases by orders of magnitude and subsurface water is commonly undersaturated with uranium. Uranium absorbed by secondary minerals, particularly by iron hydroxides and leucoxene, is its single stable species under oxidizing conditions. The impact of oxygen-bearing water leads to destruction of uranium ore. This process is realized simultaneously at different hypsometric levels even if the permeability of the medium is variable in both the lateral and vertical directions. As a result, intervals containing uranyl minerals and relics of primary uranium ore are combined in ore-bearing zones with intervals of completely dissolved uranium minerals. A wide halo of elevated uranium contents caused by sorption is always retained at the location of uranium ore entirely destroyed by weathering. Uranium ore commonly finds itself in the aeration zone due to technogenic subsidence of the groundwater table caused by open-pit mining or pumping out of water from underground mines. The capillary and film waters that interact with rocks and ores in this zone are supplemented by free water filtering along fractures when rain falls or snow is thawing. The interaction of uranium ore with capillary water results in oxidation of uraninite, accompanied by loosening of the mineral surface, formation of microfractures, and an increase in solubility with enrichment of capillary water in uranium up to 10−4 mol/l. Secondary U(VI) minerals, first of all, uranyl hydroxides and silicates, replace uraninite, and uranium undergoes local diffusion redistribution with its sorption by secondary minerals of host rocks. The influx of free water facilitates the complete dissolution of primary and secondary uranium minerals, the removal of uranium at the sites of groundwater discharge, and its redeposition under reducing conditions at a greater depth. It is evident that the conditions of the upper hydrodynamic zone and the aeration zone are unfit for long-term insulation of SNF and high-level wastes because, after the failure of containers, the leakage of radionuclides into the environment becomes inevitable.

Modelling of radiation field around spent fuel container

Operation of nuclear reactors leads to the production of spent nuclear fuel (SNF). There are two basic strategies of SNF management: ultimate disposal of SNF in geological formations and recycle or repeated utilisation of reprocessed SNF. In both options, there is an urgent necessity to study radiation properties of SNF. Information about SNF radiation properties is required at all stages of SNF management. In order to reach more effective utilisation of nuclear materials, new fuel cycles are under development based on uranium-plutonium, uranium-thorium and some other types of nuclear fuel. These promising types of nuclear fuel are characterised by quite different radiation properties at all the stages of nuclear fuel cycle (NFC) listed above. So, comparative analysis is required for radiation properties of different nuclear fuel types at different NFC stages. The results presented here were obtained from the numerical analysis of the radiation field around transport containers of different SNF types and in SNF storage. The calculations are carried out with the application of the computer code packages SCALE-4.3 and MCNP-4C. Comparison of the dose parameters obtained for different models of the transport container with experimental data allowed us to make certain conclusions about the errors of numerical results caused by the approximate geometrical description of the transport container.

The Economics of Reprocessing versus Direct Disposal of Spent Nuclear Fuel

We assess the economics of reprocessing versus direct disposal of spent fuel. The uranium price at which reprocessing spent fuel from light water reactors (LWRs) and recycling the resulting plutonium and uranium in LWRs would become economic is estimated for a range of reprocessing prices and other fuel cycle costs. The contribution of both fuel cycle options to the cost of electricity is also estimated. A similar analysis is performed to compare fast neutron reactors (FRs) with LWRs. We review available information about various fuel cycle costs, as well as the quantities of uranium likely to be recoverable at a range of future prices. We conclude that the once-through LWR fuel cycle is likely to remain significantly cheaper than recycling in either LWRs or FRs for at least the next 50 yr, even with substantial growth in nuclear power.

A Web of Interests: Stalemate on the Disposal of Spent Nuclear Fuel

The United States has adopted a policy of geological disposal to resolve the problem of permanent disposition of highly radioactive spent nuclear fuel (SNF), but has been unable to implement this policy. This study draws upon literature on policymaking and interest group representation to examine that inability. It develops and explores a conceptualization of policy stalemate to analyze the case.The study finds that the United States has arrived at a stalemate over SNF disposal policy in part because of its pluralist system of interest group representation, and in part because of the nature of the problem. The case demonstrates how, in a pluralist system, the interaction of interests over a complex and difficult policy problem can result in stalemate. Few of the involved interest groups have sufficient incentives to either implement current SNF disposal policy or to revisit the policy. Barring significant change in the nature of the problem, the balance of interests, or other factors in the policy environment, the stalemate will continue.The central question of this symposium is how institutionalized systems of interest group representation manage competing demands for environmental protection with demands for energy. In the case of SNF disposal, the United States in effect is managing these competing demands by not addressing them. SNF disposal policy does not appear to pose directly the expected tradeoff between demands for environmental protection and accessible energy, as the conflict tends to be defined primarily in environmental terms. However, the trade‐off is implicit; failure to resolve the problem will eventually impact continued access to nuclear energy.

Conceptual basis of a systems prioritization methodology for the Waste Isolation Pilot Plant

A systems prioritization methodology (SPM) is under development at Sandia National Laboratories (SNL) to provide guidance to the US Department of Energy (DOE) on experimental programs and design modifications to be supported in the development of a successful compliance certification application to the US Environmental Protection Agency (EPA) for the Waste Isolation Pilot Plant (WIPP) for the geologic disposal of transuranic (TRU) waste. The purpose of the SPM is to determine the probabilities that the implementation of different combinations of experimental programs and design modifications, referred to as activity sets, will lead to compliance with 40 CFR 191, Subparts B and C ( Environmental Radiation Protection Standards for the Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Waste) and 40 CFR 268·6 ( Petitions to Allow Land Disposal of a Waste Prohibited under Subpart C of Part 268, which implements the Resource Conservation and Recovery Act, i.e., RCRA). Appropriate tradeoffs between compliance probability, implementation cost and implementation time can then be made in the selection of the activity set to be supported in the development of a licensing application. Determination of compliance probabilities for individual activity sets involves probability spaces for (1) possible outcomes of the experimental programs, (2) uncertainty in analysis input given specific experimental outcomes and (3) possible future occurrences at the WIPP, and also models for (1) fluid flow in the vicinity of the repository, (2) radionuclide release from the repository due to flowing groundwater, (3) groundwater flow and radionuclide transport in geologic formations overlying the repository, (4) radionuclide release to the surface environment due to cuttings and spallings removal in the event of a drilling intrusion and (5) transport of RCRA contaminants in gas and brine. Descriptions are given for the conceptual structure of the SPM and the manner in which this structure determines the computational implementation of an example SPM application. Due to the sophisticated structure of the SPM and the computational demands of many of its components, the overall computational structure must be organized carefully to provide the compliance probabilities for the large number of activity sets under consideration at an acceptable computational cost. Conceptually, the determination of each compliance probability is equivalent to a large numerical integration problem.

Uncertainty and Sensitivity Analysis for Gas and Brine Migration at the Waste Isolation Pilot Plant: Fully Consolidated Shaft

Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis, stepwise regression analysis, and examination of scatterplots are used in conjunction with the BRAGFLO model to examine two-phase flow (i.e., gas and brine) at the Waste Isolation Pilot Plant, which is being developed by the US Department of Energy as a disposal facility for transuranic waste, to provide insights on factors that are potentially important in showing compliance with applicable regulations of the US Environmental Protection Agency. Specific regulations include Petitions to Allow Land Disposal of a Waste Prohibited Under Subpart C of Part 268 (40 CFR 268.6), which implements the Resource Conservation and Recovery Act and establishes maximum environmental concentrations for regulated chemicals such as volatile organic compounds (VOCs) and heavy metals, and Environmental Standards for the Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes (40 CFR 191, Subpart B), which places a probabilistic limit on allowable radioactive releases from a disposal facility over a 10,000-yr time period. The primary topics investigated are (a) gas production due to corrosion of steel, (b) gas production due to microbial degradation of cellulosics, and (c) gas migration into anhydrite marker beds in the Salado Formation,more » which is the host unit into which the waste will be emplaced. Gas production and movement is of particular importance in establishing compliance with 40 CFR 268.6 because of its influence on the movement of VOCs. Important variables identified in the analysis include (a) initial brine saturation of the waste, (b) stoichiometric terms for corrosion of steel and microbial degradation of cellulosics, and (c) gas barrier pressure in the anhydrite marker beds.« less

96/00314 Study of the structure and components of tar from Shenfu coal

After the final shutdown of Ignalina NPP Unit 1 in 2004 and Unit 2 in 2009, total amount of spent nuclear fuel (SNF) is approximately 22,000 of fuel assemblies. Radionuclide content in SNF and its characteristics are initial data for analysis of various subjects such as radiation shielding, thermal analysis and other processes that can affect the performance of SNF storage or disposal facilities’ components. For instance, ageing of the components is influenced by decay heat, neutron and gamma radiation, external environment conditions, mechanical stresses, etc. The first two factors can be defined only after SNF characteristics are investigated and irradiation levels are determined. Experimental investigation of radionuclide content and characteristics in SNF is expensive and complicated, so numerical methods are widely used. In this paper, results of numerical modelling of spent RBMK-1500 nuclear fuel characteristics and its variation with time are presented. Modelling was performed using SCALE computer code system TRITON module. Some calculation results are compared with available experimental data for RBMK-1000 spent nuclear fuel and a rather good agreement between numerical and experimental results is identified. Comparison of calculated gamma and neutron sources, activities and decay heats shows that results for RBMK-1500 fuel with various enrichments differ by 10–20%.