Specific heat capacity model for compacted bentonite buffer materials

An engineered barrier system (EBS) for the geological disposal of high-level radioactive waste (HLW) consists of a disposal canister packed with spent fuel, buffer material, backfill material, and gap-filling material. The buffer material fills the space between the canister and the near-field rock, thus serving to restrain the release of radionuclides and protect the canister from groundwater penetration. Furthermore, as significant amounts of heat energy are released from the canister to the surrounding rock, the thermal conductivity of the buffer plays an important role in maintaining the safety of the entire disposal system. Therefore, given the high levels of heat released from disposal canisters, this study measured the thermal conductivities of compacted bentonite buffers from Gyeongju under temperature variations ranging 25 to 80~90℃. There was a 5~20% increase in thermal conductivity as the temperature increased, and the temperature effect increased as the degree of saturation increased.

Investigation of the various properties of several candidate additives as buffer materials

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

Effects of graphene content and dry density on swelling pressure of graphene-modified GMZ bentonite

ABSTRACT: Thermo-hydro-mechanical (THM) processes within a bentonite buffer are important for the performance of the Engineered Barrier Systems (EBS) in a deep geological repository for nuclear waste. The THM behavior of Engineered Barrier Systems (EBS) with high-level radioactive waste cannisters have been extensively studied for temperatures up to 150 °C. Recent research has been focused on investigating the behavior of the bentonite buffer at high temperatures (150 – 200 °C) which is motivated by expanding the data and knowledge base for confidence building. This study presents the ongoing HotBENT field test at the Grimsel Test Site in Switzerland. The field test consists of 4 heaters placed in series with target temperatures ranging from 175 - 200 °C in a granite tunnel. A 2-D numerical model was developed considering a cross section around Heater 1 which was maintained at 200 °C. Coupled thermal, hydrological and mechanical processes within the barrier system are modeled using the TOUGH-FLAC simulator. The model considers material zones for bentonite, concrete floor, excavation damaged zone and granitic host rock. The simulated results were compared with field test results for temperature, relative humidity and pore water pressure and the impact of thermal and hydrological parameters on the buffer performance was evaluated. The initial modeling results provide an insight into the behavior of the bentonite buffer at high temperatures. 1. INTRODUCTION Multibarrier deep geological repositories are preferred worldwide for the disposal of high-level radioactive waste. This multibarrier system will typically consist of a metal cannister containing the waste placed within a suitable host rock. The space between the cannister and the host rock is filled with bentonite. Bentonite is chosen as a backfill material for its favorable properties including low hydraulic conductivity and high swelling potential. The Engineered Barrier System (EBS) which consists of the waste cannister and the backfill material, is subjected to elevated temperatures from radioactive decay and fluid inflow from the surrounding rock. The heat transfer through the buffer also results in thermally induced excess pore water pressures and moisture flow through diffusion. As a result of these processes, the buffer material undergoes mechanical deformation through swelling and/or shrinkage. Understanding these strongly coupled thermo-hydro-mechanical (THM) processes within the bentonite is crucial when designing a safe and efficient repository.

ABSTRACTWater vapor adsorption characteristics of buffer materials to be used for the geologic disposal of high-level radioactive waste have been investigated. Highly compacted blocks of Japanese Na bentonite were used as the buffer material. Initial dry density of the blocks was 1.8 g/cm3. Volume changes and water contents of the blocks in equilibrium with surrounding moist air were measured. The humidity of the surrounding air was kept constant, and the suction of the block at equilibrium has been determined by the humidity. The temperature of the surrounding air was varied between 20°C and 60°C.The volumetric strain was observed to vary approximately between −3% and 13% depending on the humidity of 0.6% to 95%. Some analytical expressions derived from these experiments on moisture adsorption are presented in this paper. Water vapor adsorption isotherm of the bentonite blocks was also obtained. The swelling pressures calculated from the analytical expressions are compared with literature data for this bentonite.

Concepts for geological disposal of high-level radioactive waste usually include bentonite buffer materials. Numerous studies have been performed with most usingWyoming bentonite. Gaomiaozi (GMZ) bentonite has been selected as a potential buffer/backfill material for the deep geological repository of high-level radioactive waste in China. In this context, the highly alkaline environment induced by cementitious materials in the repository is likely to alter montmorillonite, the main clay mineral in GMZ bentonite. This alteration may result in deterioration of the physical and/or chemical properties of the buffer material. To acquire quantitative data which would allow us to assess the dissolution of montmorillonite and changes in the diffusivity of hydroxide ions as well as their effects on the swelling pressure and permeability of the compacted GMZ bentonite, an experimental study was conducted under highly alkaline (NaOH solutions with various pH values were used), simulated groundwater conditions. The GMZ bentonite also contains cristobalite which may also have been dissolved. The microstructure of the compacted bentonite samples after the experiments was determined by mercury intrusion porosimetry (MIP) and scanning electron microscopy (SEM). Energy dispersive spectroscopy (EDX) was carried out to identify mineralogical changes. At pH >13, the permeability of specimens increased significantly; the swelling potential decreased with increasing pH. Furthermore, the pore volume and pore size of GMZ bentonite changed when exposed to alkaline solution, resulting in an increase in porosity and permeability. The main alteration mechanisms of compacted GMZ bentonite undergoing infiltration by highly alkaline solution are likely to be dissolution and modifications in terms of the microstructure and mineralogy.

Activation energy for diffusion of chloride ions in compacted sodium montmorillonite

Many countries rely on nuclear energy. However, ensuring safe disposal of high-level radioactive waste (HLW) generated has become an issue. Deep geological repository, the most recognized method for long-term disposal of HLW, isolates HLW from the biosphere using containers with multiple barriers buried in stable geological formations. In the early stages of storage, the corrosion behavior of the containers would be affected by heat dissipation of the waste, alkaline buffer layers, and ions from groundwater. In mid-storage, as heat dissipation decreases and temperature drops, microbial activity increases, making microbiologically influenced corrosion (MIC) a significant concern. This study examined corrosion behavior of copper, 316L stainless steel, low carbon steel, and Alloy 22 under simulated early storage conditions (80°C in aerated calcium hydroxide solution containing chloride, sulfur, and ammonium ions) using mass change and polarization measurements. Microbial corrosion tests were conducted at 50°C in anaerobic environment containing sulfate-reducing bacteria to analyze microbial corrosion and electrochemical properties using surface analysis, mass change analysis, and electrochemical measurement. The results indicated that under simulated early storage conditions, all four materials exhibited the highest corrosion trend in solution containing only chloride ions. Furthermore, nickel-based alloys showed the best corrosion resistance in all environments.

고준위 방사성 폐기물 처분시스템은 지하 500 m 심도에서 암반에 터널을 뚫어 고준위폐기물 처분용기를 넣고 주위를 완충재로 매우는 형태이다. 많은 통계 자료에 의하면 한반도에서 매년 지진이 증가하는 추세이며, 지진이 발생할 경우 지하에서 발생된 전단력에 의해 처분용기가 손상될 수 있다. 더 나아가 방사성 유해물질이 유출되어 큰 환경 문제가 유발될 수 있다. 이에 본 논문에서는 지진에 대해 안전하게 보호할 수 있는 방법으로 내진형 완충재를 개발하였다. 내진 성능에 영향을 미치는 주요인자를 분석하여 내진형 완충재를 설계하였고, ABAQUS를 이용하여 전단해석모델을 개발하여 내진형 완충재의 성능을 평가하였다. In the underground 500 m depth, the high level radioactive waste disposal system is made by boring the tunnel in the base rock and putting the high level waste disposal canister that is the surrounding form with the buffer material. According to the many statistics, it is the tendency that the earthquake increases in the Korean peninsula every year. In case that the earthquake is generated, the disposal canister in the rock mass can be broken due to the shearing force in the underground. Furthermore, a major environmental problems can be caused by the radioactive harmful substances. In this study, the earthquake-proof type buffer material was developed with the protection method safely on the earthquake. The main parameter having an effect on the earthquake-resistant performance was analyzed and the earthquake-proof type buffer material was designed. The shear analysis model was developed and the performance of the earthquake-proof type buffer material was evaluated by using ABAQUS.

Congress passed the Nuclear Waste Policy Act of 1982 (NWPA) and its amendments establishing the National policy for safely storing, transporting and disposing of spent nuclear fuel and high-level radioactive waste in a geologic repository. This legislation created the Office of Civilian Radioactive Waste Management (OCRWM) within the US Department of Energy (DOE) to develop an integrated system for the safe and efficient disposal of high-level radioactive waste. The NWPA, as amended, directs DOE to study in detail the Yucca Mountain site in Nevada as the only candidate site for the Nation`s geologic repository. In Nevada, the DOE/OCRWM Yucca Mountain Site Characterization Project Office (YMPO) manages scientific investigations to determine the suitability of Yucca Mountain for effectively isolating radioactive waste from the public and the environment. This Information Services Directory is intended to facilitate dissemination of information. The Directory is produced by the Education and Information Division of OCRWM`s Office of External Relations and will be updated periodically. This is the third such update since its issuance in August 1986. It is a reference document that lists the sources of program information available to States, Indian Tribes and the public.

A new rock mass classification system QHLW for high-level radioactive waste disposal

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technologies (OFCT) has established the Used Fuel Disposition Campaign (UFDC) to conduct research and development (R&D) activities related to storage, transportation and disposal of used nuclear fuel (UNF) and high level radioactive waste (HLW). The U.S. has, in accordance with the U.S. Nuclear Waste Policy Act (as amended), focused efforts for the past twenty-plus years on disposing of UNF and HLW in a geologic repository at Yucca Mountain, Nevada. The recent decision by the U.S. DOE to no longer pursue the development of that repository has necessitated investigating alternative concepts for the disposal of UNF and HLW that exists today and that could be generated under future fuel cycles. The disposal of UNF and HLW in a range of geologic media has been investigated internationally. Considerable progress has been made by in the U.S and other nations, but gaps in knowledge still exist. The U.S. national laboratories have participated in these programs and have conducted R&D related to these issues to a limited extent. However, a comprehensive R&D program investigating a variety of storage, geologic media, and disposal concepts has not been a part of the U.S. waste management program since the mid 1980s because of its focus on the Yucca Mountain site. Such a comprehensive R&D program is being developed and executed in the UFDC using a systematic approach to identify potential R&D opportunities. This paper describes the process used by the UFDC to identify and prioritize R&D opportunities. The U.S. DOE has cooperated and collaborated with other countries in many different “arenas” including the Nuclear Energy Agency (NEA) within the Organisation for Economic Co-operation and Development (OECD), the International Atomic Energy Agency (IAEA), and through bilateral agreements with other countries. These international activities benefited the DOE through the acquisition and exchange of information, database development, and peer reviews by experts from other countries. Recognizing that programs in other countries have made significant advances in understanding a wide range of geologic environments, the UFDC has developed a strategy for continued, and expanded, international collaboration. This paper also describes this strategy.

A set of thermo‐hydro‐mechanical coupling control equations was established, and the healing process of the joints between Gaomiaozi bentonite (GMZ01) buffer material blocks and the influence of the joint parameters were numerically simulated. The calculations consider the effect of joints on solute migration, the permeability and thermal conductivity of the buffer material, and the evolution of the healing effect. The effects of the joint design parameters, including the type, number, width, splicing form, and average dry density of the joint, are investigated. Studies show that, under an external water head, the joints will become hydraulic priority channels due to their higher permeability, which will shorten the saturation time of the blocks. As the bentonite gradually saturates, the swelling force compresses the joint material. This action improves the overall uniformity of the buffer material and reduces the priority channel effect. Meanwhile, the final average permeability and diffusion coefficient of the buffer material are found to mainly depend on the average dry density of the buffer material. The higher the average dry density of the buffer material is, the lower the final average permeability and diffusion coefficient are, whereas the distribution of joints and the block splicing are less affected by the average dry density of the buffer material. The findings of this study can provide a reference for the design of bentonite buffer material blocks in the repository.

Salt formations have long been recognized as a highly favorable host rock for the final disposal of high-level radioactive waste (HLW) in deep geological repositories. Their unique properties, including exceptional impermeability, self-healing capabilities, and thermal conductivity, make them a reliable natural barrier for the deep disposal of radioactive waste. This report focuses on the development and application of a methodology for assessing the integrity and per formance of the Engineered Barrier System (EBS) within salt-based repositories, a critical component of the multi-barrier system ensuring safe radioactive waste disposal.