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DOE OSTI2016

Developing a New National Programme and Implementation Plan for the Management and Disposal of Spent Fuel and Radioactive Wastes in Romania - 16391

The European 'Waste Directive' (2011/70/Euratom) required each Member State to report by August 2015 on their National Programme for management of spent fuel and radioactive wastes. In Romania, the Nuclear Agency (ANDR) was responsible for preparing the National Programme report, and this was done in a step-wise manner that involved the regulators and waste producers. The starting point for the process was an assessment of existing policy against the evolving situation in the nuclear sector in the country, most notably the decision to progress with the construction of two new CANDU reactors. The final agreed National Programme report submitted to the European Commission describes the activities currently undertaken to manage spent fuel and radioactive wastes in Romania. The report also sets out, at a high-level, the additional activities that will need to be taken in the future (and an indicative timeline), including those steps that will be required to implement a new surface repository for short-lived wastes, and a geological repository for spent fuel and long-lived wastes. Additional, on-going work is intended to lead to a revision of Government policy and the development of a more detailed implementation strategy, underpinned by a programme of R and D and robust financing plans. (authors)

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DOE OSTI2016

Perspective of Dry Interim Storage of spent nuclear fuel (SNF) in Germany - 16504

According to article 6 of the German Atomic Energy Act, storage of nuclear fuel requires a license granted by the competent authority in this field, which is the Federal Office of Radiation Protection (BfS). In 1995 a license to store spent nuclear fuel (SNF) and vitrified high level waste (HAW) for the interim storage facility in Gorleben was granted. Dry Interim storage at the NPP sites in its current form started in 2002 in the interim storage facility next to the NPP Lingen. Since this time each NPP erected its own storage facility and three central storage facilities have been built. All of these facilities use dry storage in metallic dual purpose casks (DPCs). The actual storage licenses for SNF and the HAW (which is stored in Gorleben and Rubenow) are limited to 40 years after closing the transport and storage cask. The first license to expire will be the storage license for the interim storage facility in Gorleben in the year 2035. During the last two years in Germany preparatory consultations to restart the process towards finding a repository took place, organised by a commission. This commission consists out of members of the scientific community, religious parties, social communities and environmental organizations as well as members of federal states governments and members of the German parliament (Bundestag). Recently the commission decided to extend its working period until summer 2016. Taking this background into account the BfS plans to intensify its research into extended interim storage with respect to potential material degradation leading to interference with safety functions (e.g. neutron shielding material). This includes topics like ageing of gasket materials which are already studied in ongoing experiments (e.g. at the Federal Institute for Materials Research and Testing) as well as studies of long term relaxation processes. Because of the need for transportation towards a repository another important field is the behaviour of the inventory under mechanical loads. For SNF, first calculations showed that because of the lower decay heat load after 60 years due to hydride reorientation the cladding might become much less ductile. (authors)

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DOE OSTI2016

Future German R and D on HLW Disposal - 16519

In 2009, after the General elections, the new German Government showed a relatively positive attitude towards a prolonged use of nuclear power. Power plant shutdown was planned for the year 2032. Nuclear power was considered as a sort of transition energy technology until the energy supply would take place by the use of renewable energy sources. This was in line with the legal framework and the political strategy of the energy turnaround. Then, 17 nuclear power plants (NPP) were operational. The Fukushima incident heralded the sound of drastic changes in the German energy policy: Besides an intermediate stop of operation of reactors until a so-called stress test was accomplished, the general political decision was to phase-out nuclear earlier than planned and to implement faster the decisions regarding the energy transition policy. The shut-down of the last German nuclear power plant will occur in 2022. Accordingly, all the necessary legal changes were made. A consensus about the appropriate waste management strategy for heat generating waste (spent nuclear fuel (SNF) and vitrified waste (HLW)) was still pending at this time. Public and political discussions on this subject continued as in the years before. In July 2013, after intensive discussions, most of the parties represented in the German parliament agreed to eventually tackle the challenge of finding a repository by a consensual, transparent and science-based approach. This was to be codified in a law. In July 2013, the Site Selection Act passed the parliament. This law should pave the way for a new holistic process that addresses social, political, and scientific aspects of siting and the site selection process. The basics for this process should be provided by the 'Commission Storage of Radioactive Waste Materials' which was established according this law. Against this background discussions showed that scientific-technical R and D activities should be mirrored accordingly and R and D has to deal both with the future challenges of radioactive waste HLW disposal and related aspects. Considering this the Federal Ministry for Economic Affairs and Energy (BMWi), which is responsible for non-site specific R and D, created the new Research Concept 'Research for the management of radioactive waste', being fundamental and the frame for non-site specific R and D projects funded by BMWi. Key changes - mirroring the new requirements and different aspects - touch the issues: host rock research, i.e. all potential host rocks available in Germany are equally addressed and there is no prioritization, consideration of consequences of prolonged storage at NPP sites, alternative waste management strategies compared to direct disposal in a mined repository, and involvement of socio-technical aspects.This paper intends to highlight the key changes, its consequences, and will give a brief overview on first initiated R and D activities. (authors)

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DOE OSTI2016

Millennial Time Capsules as a Promising Means for Preserving Records for Future Generations - 16542

When we deal with preservation of Records, Knowledge and Memory (RK and M) for any long-term, non-inspected facility such a deep geological repository, there is no single technical or cultural provision that can be relied upon to do alone the preservation job 100%. Rather we should increase our chances by implementing a combination of approaches based on different components that provide redundancy and/or pointers to one another. Time capsules are no exception. However, they seem well suited to support national archives and other preservation elements, in order to preserve RK and M as long as possible or interesting. Time capsules are a ready to go, workable concept, with many examples of implementation of large-size, millennial time capsules at small depths based on science and lessons to be learned. A proposal is also made for considering and developing small-size time capsules placed strategically deep underground at repository level. (authors)

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DOE OSTI2016

Views on Long-Term Safety of Deep Borehole Disposal - 16657

The Swedish Nuclear Fuel and Waste Management Co., SKB, has since the end of the 1980s evaluated Deep Borehole Disposal (DBD), of spent nuclear fuel (SNF). This has been done in parallel with the development of a mined repository concept based on the KBS-3V method. The most recent evaluation involved a broad comparison between the KBS-3V method and DBD. The comparison highlights differences between the two methods in the whole chain of the SNF handling including premises for siting, licensing, repository construction, safety during the operational phase, post-closure safety, physical protection, nuclear safeguards, retrieval of the deposited SNF, time planning, need for development, costs and nontechnical project risks. The ambition has been to make an as objective comparison between the two methods as possible, bearing in mind that there are big differences between the two concepts in terms of technical maturity as well as in terms of both quantity and quality of available data. While the primary safety function of the KBS-3V method is long-term containment in a corrosion resistant copper canister, the safety of DBD relies on stagnant groundwater at depth. It is argued in the aforementioned study that this groundwater stagnancy may be jeopardised because of heat and gas evolution in the disposal boreholes, and that deformation of the boreholes make efficient sealing of the deposition boreholes difficult. SKB has taken the position that the DBD method is associated with too many question marks to make it interesting as an alternative method in the Swedish programme. (authors)

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DOE OSTI2018

Operational Safety and Post-Closure Safety: Investigation of their Interactions by a FEP Analysis - 18040

The consideration of operational safety for a repository for high-level waste is a challenge, since it has to integrate the dimension of long-term safety. The operator has to demonstrate on the one hand the safety of the workers in the mine and of men and environment during operation. On the other hand a proper operation of the mine has to guarantee that the state of the facility at the time of closure will be compatible with the requirements for long-term safety. To achieve this, a pre-closure safety analysis should include (a) a general description of the structures, systems, components, equipment, and process activities during the operational phase, (b) identification and a systematic analysis of naturally occurring and human-induced hazards and (c) identification and documentation of the strong relationships between operational safety and post-closure safety. In a German R and D project, a FEP catalogue for the operational phase was developed based on the national disposal concepts for clay and salt. The FEP catalogue contains analyses of the processes and events that affect the components (features) of the repository system in its entirety. The FEP catalogue was applied to identify hazards for the operational phase and to document the impacts of these FEP (including the identified hazards) on post-closure safety. (authors)

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DOE OSTI2018

Building Sustainable and Efficient Markers to Bridge Ten Millennia: The Understanding Issues-18058

Markers of nuclear waste deposits that can be understood through ten millennia have been proposed for preventing from inadvertent human intrusion. It is postulated that currently designed waste repository markers are likely to fail, and an alternative system of markers is proposed based on a praxeological device, i.e., a language-learning by- action device. The key challenge is how should we represent radioactivity in waste repositories in such a way that it is understood by anyone who encounters such markers? Namely, for an adequate interpretation, two issues have to be addressed: to accommodate differing world views, and inevitable changes in languages. What is required is an integrated system that allows us to communicate a complex and abstract concept to all people at all times. That is the goal of our praxeological device, a sign that links a person's actions (praxis) in a particular context to a communication system about those actions and that particular context. That means anyone who experiences such a praxeological device will understand the actions that s/he is about to perform will help her/him to learn the iconic language s/he is reading. (authors)

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DOE OSTI2018

Deposition Behavior of Supersaturated Silicic Acid onto Illite as an Altered Backfilling Material - 18114

The pH gradient, which is generated by cement-based materials, causes the deposition of silicic acid in the backfilled tunnels in the repository. This deposition behavior blocks the flow-path, whence it also acts effectively for the retardation of radionuclides besides the similar effect with swelling property. Such deposition effect is expected to be maintained in the backfilling material even after illitization has progressed. The aim of this study is to evaluate the deposition rate constant, k [m/s], of supersaturated silicic acid on illite under some different conditions, considering its temperature dependence. In the experiments, deposition behavior was observed during 6 hours. After adding illite powder into the supersaturated silicic acid solution (pH>10), the pH was set to 8 in order to adjust the solution supersaturated. As a result of these experiments, the values of k were several 10-10 m/s in the temperature range of 288 K to 323 K. Several previous studies with some experiments and numerical simulations showed that the values of k for various minerals, e.g. Ca-type bentonite and amorphous silica, were in the range of 10-11 to 10-10 m/s, and that those values are large enough to clog micro flow-paths. These means that illite shows the clogging effect caused by the deposition of supersaturated silicic acid as in other silicate minerals. In addition, the deposition amount of supersaturated silicic acid increased as the temperature was raised. Therefore, the deposition reaction could progress rapidly under the condition of relatively high temperature caused by the geothermal heat and the waste heat in the repository. (authors)

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DOE OSTI2018

The Retrieval project Asse II - Lessons Learned in Radiological Protection - 18263

From 1967 to 1978 in total 125,787 drums and packages with low- and intermediate level waste were emplaced in the Asse II mine. The disposal operation ended in 1978 due to a revision of the Atomic Energy Act. For the disposal of radioactive waste, a plan approval procedure would have been mandatory due to the Atomic Energy Act, but the mining law was the legal basis for operating the Asse II mine, further on. A concept for a closure of the mine was not existing at that time. On January 1, 2009, the Federal Office for Radiation Protection (Bundesamt fuer Strahlenschutz BfS) became the new license-holder and operator of the Asse II mine under the Atomic Energy Act. The Act on the Organizational Restructuring in the Field of Disposal in Germany became effective on July 30, 2016. The act provides that a federally owned company, the Bundesgesellschaft fuer Endlagerung (BGE mbH, Federal Company for Radioactive Waste Disposal) is formed to carry out the operative tasks of searching for a High Level Waste (HLW) repository, the construction and operation of repositories (Konrad, Morsleben) and of the Asse II mine. Since April 25, 2017 BGE is in charge of the operative tasks that have so far been carried out by BfS. Up to now, the long-term safety of the Asse II mine has not been demonstrated in the case that all the waste will remain inside the mine. Further problems arise from a lack of host rock stability, the incomplete backfilling of the mining areas that have not been used for the emplacement of radioactive waste, and a continuous brine inflow into the mine of approx.12 m{sup 3} per day. For these reasons, according to chap. 57b of the Atomic Energy Act, the Asse II mine is to be closed without delay after retrieving the emplaced nuclear waste. Besides the recent organizational changes in Germany and the special problems giving rise to retrieve the waste, in this paper the operational experiences since 2009 on the way from mining law to atomic law will be described. This applies to the radiation protection regime, the release and clearance procedure, and some measures to boost efficiency. (authors)

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DOE OSTI2018

Updated Declaration of Legacy Waste Based on Re-evaluated Records - 18283

Kerntechnische Entsorgung Karlsruhe (KTE) has the assignment to decommission and dispose of the nuclear installations on the Karlsruhe site. The entire inventory of these installations as well as the existing waste materials stored in the interim storage of the central waste management section (formerly central decontamination department / HDB) is planned to be sent to the repository Konrad commissioned for the acceptance of low level waste and intermediate level waste. Therefore it is treated according to the respective acceptance criteria. KTE has approximately 45.000 m{sup 3} of legacy waste, which was produced prior to the first issue of these acceptance criteria [1] and hence has to be re-qualified. Legacy waste generally has a radiological declaration, which is not sufficient to meet the criteria applied by the regulators and independent experts today. In order to achieve a modern declaration, additional information such as acquisition records, operational reports, calculated activation, old and new analyses and burn-up calculations as well as knowledge of the physical and chemical boundary conditions of defined material streams is needed. At KTE the waste was treated in batches according to accepted methods documented in the database system KADABRA. For each batch, records regarding the waste type, activity and mass are available for reevaluation. Each batch record consists of information on the raw waste treated in the batch, on the batch itself and on the resulting interim products or waste products. On the basis of this existing or additionally acquired knowledge a declaration concept matching the existing information and the requirements of the repository can be derived and discussed with the independent expert and the regulator. (author)

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DOE OSTI2018

Management of Metals from Decommissioning of Nuclear Facilities - 18411

The EDF group companies within the Cyclife platform have for decades performed treatment and recycling of metals. With different approaches but all considered international Good Practice, as valuable materials are recovered for safe recycling and volumes in the repositories for radioactive waste are saved. For an efficient decommissioning project, a well-defined waste management process is one of the critical paths, not least when it comes to routes for clearance and recycling of materials. It is easy to underestimate the complexity and bottlenecks caused by shortage of logistic space. In order to mitigate these risks, activities such as strengthening categorization, quantification and adopting different routes based on the material properties should be implemented. By categorization of the of waste and materials considered/classified as contaminated or potentially contaminated it is obvious that the management of the metals is a key parameter in nuclear decommissioning. Analyses performed by the EDF companies shows that a majority of the materials, classified as contaminated, from a BWR/PWR plant are metals. Experience gives that a high percentage of these metals from decommissioning can be subject to recycling either directly upon completion of the clearance process or after certain treatment. Non-technical key parameters like the regulatory requirements on applying the waste hierarchy, the impact on repository programs as well as, not at least, the regulatory and public acceptance of recycling of materials from the nuclear industry are, together with costs essential factors when forming the routing of metals and other waste materials from a decommissioning project. Today the EDF group companies are positioned to provide a wide range of services for management of the contaminated metals from decommissioning projects, spanning from industrial management of metals which directly will be subject general clearance after verification measurements up to decontamination and melting of selected reactor internals for recycling to public domain or within the nuclear industry. In this paper the typical waste material composition for BWR and PWR plants, the different ways to treat/manage the radioactive and potentially radioactive metals including a comparison of the different approach commonly used in a technical, logistical and economical perspectives. A few success stories will be resented as well as lessons learned. (authors)

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DOE OSTI2018

Enhanced Evaluation Criteria to Meet the WIPP Waste Acceptance Criteria - 18655

The Waste Isolation Pilot Plant (WIPP) is a deep geologic repository for the disposal of Transuranic (TRU) waste and the only such disposal facility of its kind. TRU waste is radioactive mixed waste generated by the U. S. Department of Energy in support of National Defense. In 2014, the WIPP facility was shut down when two separate accidents occurred in the repository. The two accidents were a fire on a large underground haul truck that prompted evacuation and filled the mine with smoke and soot, and a radiation release that contaminated the exhaust ventilation and disposal areas of the mine. DOE performed two accident investigations in 2014 that resulted in many corrective actions. Some of the corrective actions resulted in significant changes in the WIPP Waste Acceptance Criteria (WAC) and are required to be applied to all TRU waste upon the resumption of shipments to WIPP. WIPP re-opened and began taking shipments again in 2017 and a large backlog of over 25,000 TRU waste containers that were certified prior to the new WAC have been undergoing re-evaluation since the re-opening. This paper will focus on the new WAC criteria that TRU waste must meet above-and-beyond previous WAC criteria and the implementation of new processes for approval. The primary area for these changes is in the area of waste characterization using Acceptable Knowledge (AK) and new detailed analyses required for chemical compatibility. The accident investigation revealed that the radiation release was due to a thermal runaway and bursting of the waste container due to inadvertent mixtures of nitrates and combustibles resulting in ignitable waste forms. In other words, the TRU waste container had incompatible contents that went undiscovered and created a hazard characteristic that was prohibited at WIPP. Through detailed chemical compatibility evaluations and new oversight, the DOE should be able to prevent a similar accident (see Figure 1). The new chemical compatibility evaluations and evaluations of oxidizing chemicals combined with combustibles are complex and detailed. Chemical compatibility is not a new requirement, but a much more detailed evaluation is a new requirement. And, obtaining historical information about the generation of the final waste form for legacy packaged waste can be very challenging. Specifically, determining possible amounts of chemicals in the waste can be extremely challenging. This paper does not explore in depth, the details of oxidizing chemicals. Other papers presented will discuss oxidizing chemicals. (author)

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DOE OSTI2016

Methods to Capture the Tritium Released by Recovery/Recycle of Zirconium from Used Fuel Cladding: A Review

After uranium, the second largest mass contained in most used nuclear fuel (UNF) is the zirconium-based Zircaloy{sup R} cladding in the fuel cladding which accounts for about 25% of the mass. In current practice, the cladding requires disposal in a geologic repository. Zirconium may be recovered from the Zircaloy cladding of UNF for recycle. Alternatively, it may be decontaminated to recategorize it as low-level waste. Either approach reduces the quantities of high-level waste destined for a geologic repository, thereby reducing the overall cost of recycling UNF. Process development studies are under way in the United States and Republic of Korea based on dry chlorination using a mixture of chlorine (Cl{sub 2}) and an inert carrier such as argon (Ar) to convert the zirconium in the alloy, which is greater than 98 wt% of the alloy mass, to zirconium tetrachloride (ZrCl{sub 4}). The ZrCl{sub 4} is volatile and is easily separated from the non-volatile components. The residues include non-volatile chlorides such as CrCl{sub 3}, oxychlorides and oxides. The product ZrCl{sub 4} is recovered as a solid salt by condensation when cooled to less than about 300 deg. C. Feasibility tests with actual UNF cladding have demonstrated that the recovered zirconium chloride salt is sufficiently separated from fuel and fission product radionuclides that it emanates no significant dose and can be handled without shielding. Several methods for capturing tritium released from cladding during the chlorination process have been identified. A detailed description of the methodologies, advantages, drawbacks, and conceptual flowsheets will be presented. (authors)

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DOE OSTI2016

Fuel Cycle Study to Developing a Viable Long Term Nuclear Power Scenario for the UK

Within the UK a closed fuel cycle has been in operation for decades but with the introduction of a new build programme to create 16 GWe capacity using LWR technology and with the closure of the UK's reprocessing industry, an open fuel cycle will be implemented beyond 2020. However, it is expected that 16 GWe nuclear capacity will not be enough for the UK to curb greenhouse gas emissions through increased demand for low carbon electricity, and tackle energy security. As such further nuclear expansion scenarios are being considered between 40 and 75 GWe by 2050. This study evaluates a viable 40 GWe UK nuclear scenario, evaluating the extent to which key drivers can impact a future UK nuclear energy strategy, including; the ability to sustain a relatively high nuclear capacity, fresh uranium availability, accumulation of spent fuel at decommissioning nuclear reactor sites and capacity requirements of future GDF host sites. The thermal reactor open fuel cycle is the de-facto standard around the world for many reasons including the reliance on extremely mature technology, significant operational experience and most importantly a relatively low cost. However, the premise that a thermal reactor open fuel cycle is the correct one to adopt long term is dependent on the notion that large quantities of UO{sub 2} spent fuel can eventually be disposed of in a geological disposal facility and the future availability of fresh uranium. The total amount of available uranium is unknown and if extraction from seawater proves to be economic and practical then any future concerns regarding supply disappear. Spent fuel disposability on the other hand has never been performed on a commercial scale and there is some uncertainty when and whether this can be achieved. This paper will follow from previous work already presented in 2013 [REF] which examined UK fuel cycles involving the transition to a closed fast reactor fleet by the end of this century. The revised scenarios now assume fast reactors are required later in the scenario, in-line with other countries. In addition, the new study has used physics methods originally developed by SKB to assess the physical size of the repository required to dispose of all intermediate and high level waste whilst ensuring temperature limits imposed by the engineered barriers and surrounding geology are adhered to. The disposition of less Am-241 in the repository will no doubt lead to a reduction in GDF footprint. However, this may not be important to GDF design and economics. The purpose of this paper is to also critically review the importance of and put into perspective a smaller GDF footprint and lower fissile inventory. (authors)

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DOE OSTI2019

Scenario Development: Proposal of a Methodology for Different Host Rocks and Disposal Concepts - 19007

Scenario development is an essential part of the assessment of the long-term safety of a repository for radioactive waste. The new act on site selection in Germany in 2017 requires that salt, clay and crystalline rock are studied as possible sites for repositories. So far, a comprehensive methodology for scenario development has been developed only for a site in salt rock in Germany within the Preliminary Safety Analysis (VSG). The question arose whether this methodology is transferable to other host rocks and disposal concepts and which modifications may be necessary. The transferability was shown in the project ANSICHT for clay rock sites while introducing some modifications of the methodology. This project studied the transferability of crystalline rock sites, however, there is no suitable safety and disposal concept or even a generic site yet. To study the transferability, initial barriers and FEP for different concepts were selected and the first steps of the methodology for crystalline rock were applied and discussed. It was concluded that the definition of barriers as containment providing rock zone (CPRZ) is difficult but important for the application of the methodology. Concepts for crystalline rock can foresee a CPRZ within the host rock or above. If the concept is using engineered barriers only and no CPRZ, the number of initial barriers is low. This will subsequently lead to a more complex scenario development since long FEP chains have to be considered. Based on the discussion so far, the transferability of the methodology seems to be feasible in general. (authors)

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DOE OSTI2019

French Geological Disposal Project: From Siting to Permitting - 19018

The future 'Cigeo' geological repository will be located in eastern France, at the border between the Meuse and Haute-Marne. Research has been carried out in this area since the 1990's, leading to the identification of a zone within the Paris basin whose geological characteristics make it suitable for the deep geological disposal of high- and intermediate-level long-lived radioactive waste, as well as a smaller zone of approximately 250 km{sup 2} (proposed in 2005) within which the characteristics of the host layer are very similar to those observed in the Underground Research Laboratory (URL). In 2010, based on additional surveys, a zone of approximately 30 km{sup 2} called ZIRA, in which the industrial underground facility would be constructed was defined. The repository structures within the ZIRA will be located at a depth of 525 metres. Half the surface area of the ZIRA is located in a wooded area, the other half being situated in grassland or farmland. All the subsequent design studies (preliminary and detailed) of Cigeo as well as the preparation of the licensing files (to be submitted in 2020) were conducted on the basis on this precise siting. (authors)

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DOE OSTI2019

Decommissioning, Decontamination, and Disposal of LLW-LL, and ILW-LL at a Former Rare Earths Processing Facility in Malaysia - 19087

Geosyntec Consultants, Inc. (Geosyntec) was the design-build Contractor for a project involving the remediation and disposal of low level (LLW-LL and ILW-LL) NORM and TENORM wastes at a former rare earth processing facility in Malaysia. These sites were associated with a plant that extracted rare earth elements from local tin ores. Because of the naturally occurring radioactivity contained in the raw materials, these productions resulted in by-products and residues containing LLW-LL and ILW-LL thorium hydroxide and uranium hydroxide. The total mass of uranium (∼120 tons) and thorium (∼1,200 tons) managed in a single repository required the project design incorporate requirements related to inventory, security and safeguards around nuclear precursor materials. The requirement to consider a performance over a period of more than 10,000 years led to significant challenges in design and modeling of the disposal facility as this disposal facility was purposely designed and permitted as a final disposal for the LLW-LL and ILW-LL. The first phase of the project was the decommissioning, decontamination, and demolition of the 14- acre former rare earth processing plant and excavation of contaminated soils, construction of an engineered cell at the repository for waste disposal, and transportation and disposal of the waste materials into the cell. Final status surveys were carried out to allow the site to be successfully released by the Malaysian regulatory authorities as a clean site for re-development as common industrial use. Over 11,000 loads of material, including approximately 75,000 cubic meters of excavated soils and plant debris, were transported in specially designed steel transport boxes on public roads for disposal in the engineered cell. The second phase of the project included the design and construction of a second engineered cell, enhanced with protective barrier systems and a final cover system constructed of natural materials. Ancillary facilities such as substations, offices, processing buildings, and wastewater treatment plants were constructed to support retrieval and processing of the materials from more than 87,000 individual drums using a semi-automated process. The drums, waste packages, and other materials in the storage facility were systematically extracted, sorted, inventoried, solidified using Portland cement, transported, and disposed in the cell. All temporary facilities were decontaminated, demolished, and disposed in the cell. Radiological inventory and radiation safety programs were implemented to quantify the waste activity and to control exposure to workers and the public during the project. The total mass of radioactive contaminated waste disposed in the two projects is nearly 200,000 tons. The projects were carried out under a high level of scrutiny from Malaysian and international regulatory bodies. Between 2014 and 2017, performance monitoring and modeling activities were carried out to demonstrate that the disposal facility was performing in accordance with design requirements established by regulatory agencies. (authors)

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DOE OSTI2019

Estimation of Deposition Behavior of Supersaturated Silicic Acid around Geological Disposal Facility - 19137

For safety evaluation about the geological disposal system, the leakage of radionuclides with groundwater flow is estimated by considering the groundwater flow through the manmade and natural barriers. In such an estimate, we should also consider the influence of cement materials used for the construction of the disposal facility system. Cement material including high alkalinity alters groundwater condition up to 13 in pH and dissolves numerous silicate materials, which release silicic acid to groundwater. Since this groundwater mixes with natural groundwater downstream, supersaturated silicic acid will be produced with the decrease of pH to 8 around the repository system. Thus, in this study, the behavior of supersaturated silicic acid in a flow system was examined by using the column packed with amorphous silica particles under the condition of constant hydraulic head. The packed length is 20 mm or 40 mm, and the supersaturated silicic acid solution was continuously injected into the column. Here, the supersaturated silicic acid was adjusted by mixing with distilled water of pH 8 just before the solution was injected into the column. In the results, the supersaturated silicic acid quickly deposited in the packed column, and colloidal silicic acid was not detected at the outlet of the column. The deposition amount of supersaturated silicic acid in the column decreased with time, and became larger in the longer column. Besides, the flow rate decreased with the deposition of supersaturated silicic acid. Thus, this study monitored the decrease rate of permeability with the supply amount of silicic acid. In the results, the change rate did not depend on packed length. This suggested that even with a short packed length, by considering the amount of precipitation of supersaturated silicic acid, it is possible to estimate the change in permeability similarly to the case of a long packed length. Such a decrease effect in permeability may contribute to retarding the migration of radionuclides around the repository. (authors)

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