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DOE OSTI2018

Drop Testing of a Container for the Storage, Transport and Disposal of Intermediate Level Waste - 18078

Impact tests were performed, as part of a corresponding container's substantiation, during design development of a shielded container proposed for use at Sellafield for waste retrieval from the First Generation Magnox Storage Pond (FGMSP) in the UK. The mechanical test program comprised a 9 m free drop test onto an unyielding target in a container long lid edge down orientation at ambient conditions. Further, a 0,5 m free drop test onto a punch target was performed. Here, the container was orientated with the lid downwards in a way that the punch, a mild steel bar, impacts a filter lid. The test specimen was instrumented with strain gauges and accelerometers for the drop tests. Transient strains at selected points of the inner and outer container walls, at the shielding lid, as well as at the lid bolts were measured during the container's impact. Furthermore, decelerations of the container body, container lid, and the skip were measured. The complex geometrical changes of the container due to impact were determined by optical 3d- deformation measures using the projected fringes method in combination with multi-image photogrammetry. This paper summarizes the performance of the drop tests and various drop test results in context with the design development as well as aspects regarding the associated Finite Element (FE) analyses and post-test evaluation. (authors)

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DOE OSTI2019

Factors Affecting the Service-life of HDPE Geomembranes in an LLW Disposal Facility - 19320

The assessment of a suitable maximum allowable strain (MAS) for high density polyethylene (HDPE) geomembranes (GMBs) considered suitable for a proposed low level radioactive waste engineered containment mound with a 550-year design-life is described in the context of the available literature and data in 2018. Consideration is given to both the material properties and the factors contributing to the development of sustained tensile strain in the GMB. The assessment is based on the service-life of the GMB eventually being reached due to stress cracking after depletion of the protective antioxidants to a residual value. With respect to the selection of a GMB with a suitable stress crack resistance (SCR), it is noted that morphological changes can result in a substantial reduction in SCR with time, from the initial just manufactured value (SCRo) to a more representative value (SCRm), as the material tends to a thermodynamically more stable post-manufactured state. Thus, SCRm should be considered more representative of the true SCR of the material than the value obtained from the off-roll material just after manufacture. SCRm of can be estimated by aging the candidate GMBs in liquid at 55 deg. C for 270 days. Considering the potential implications of failure of the textured GMB to be used at this facility, and the potential for repair and replacement if it does fail, it is recommended that a maximum allowable strain (MAS), from all sources, of 3%, 4% and 5% be adopted on the base, side slopes, and cover GMBs, respectively. These maximum allowable strain values were selected considering the nature of the waste to be contained and the design-life that is required; they may be too conservative for some other applications with a much shorter design-life or lower consequence of premature failure. It is also indicated that the most critical locations on the GMB are often the welded seams where GMB panels are joined or repaired. Welds act a stress-raisers (increasing the stress and strain the GMB compared to that on either side of the weld). To minimize the need for patching areas damaged by the removal of samples for destructive tests, it is recommended that destructive tests on welds should generally be restricted to the minimum possible. It is also recommended that the German limits on welded GMB thickness be adopted in construction. Since strains can be significantly increased by failure to construct in accordance with the design drawings and specifications, it is recommended that there be Construction Quality Assurance (CQA) by very experienced personnel with a good knowledge of compacted clay liners, Geosynthetic Clay Liners (GCLs), and GMBs and the facility design, and that there be an experienced individual watching the construction work at all times prior to completion of the liner system. (authors)

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DOE OSTI2018

The Legislative and Regulatory Histories Related to Greater-Than-Class-C Low-Level Radioactive Waste - 18602

The history surrounding the legislative and regulatory developments regarding Greater-Than-Class C Low Level Radioactive Waste (GTCC LLW) is both interesting and complex. On October 25, 1978, the U.S. Nuclear Regulatory Commission (NRC) published an Advanced Notice of Proposed Rulemaking pertaining to the development of regulations for the land disposal of LLW. This regulation was one of the first undertaken by NRC since its inception pursuant to the Energy Reorganization Act of 1974. Given the problematic challenges that many States were facing with releases of radioactivity from waste disposal facilities at the time, the technical bases supporting this rulemaking were both forward-thinking and comprehensive. The rule included a waste classification system that was tiered in four waste classes: Class A, B, C and GTCC LLW. On December 27, 1982, the NRC finalized the requirements for disposing of these classes of radioactive waste in Title 10 of the Code of Federal Regulations, Part 61 (10 CFR 61). The NRC established the waste classification system to protect both members of the public and individuals that may someday inadvertently intrude in the waste placed in the disposal facility after expiration of the institutional control period. This waste classification was developed to ensure that radiation doses to members of the public and an inadvertent intruder would not exceed 0.25 mSv y{sup -1} (25 mrem y{sup -1}) or 5.0 mSv y{sup -1} (500 mrem y{sup -1}), respectively. At the time the regulation was promulgated, GTCC LLW was considered generally not suitable for near surface disposal. It was recognized that disposal of GTCC LLW would require disposal technologies having greater confinement capacity or protection than 'normal' near-surface disposal. In 1980 Congress enacted the Low-Level Waste Policy Act of 1980, as amended in 1985 (LLWPAA of 1985) that assigned responsibility of disposing of GTCC LLW generated in the commercial sector or owned or generated by the federal government to the U.S. Department of Energy (DOE). At the time States were having difficulties in developing disposal facilities within the time constraints mandated in the LLWPA of 1980. Therefore, Congress believed that if DOE was assigned responsibility of developing disposal facilities for GTCC LLW, then the States would be more likely to develop new LLW disposal facilities. Congress also assigned the NRC with the responsibility to license the disposal of GTCC LLW generated or owned by the federal government or by the commercial sector. Lastly, Congress changed the definition of LLW that had been defined in the LLWPA of 1980 removing the 'transuranic' exclusionary language. Four years later, on May 25, 1989, the NRC promulgated its regulation needed to implement the mandates specified in the LLWPAA of 1985. The NRC revised 10 CFR 61.55 specifying that the NRC was responsible for licensing the disposal of GTCC LLW. However, it inadvertently did not revise the definition of LLW needed to remove the 'transuranic' exclusionary language in this rulemaking consistent with the LLWPAA of 1985. This seemingly inconsequential omission was not discovered until a Petition for Rulemaking was filed in Texas in July 2014. Moreover, this omission has effectively orphaned waste with certain transuranic radionuclides with concentrations exceeding 100 nanocuries per gram (100 nCi/g). In December 2015, the NRC Commissioners directed its staff to move forward with a rulemaking to resolve this issue and to perform a regulatory analysis to determine whether or not an Agreement State can license the disposal of GTCC LLW pursuant to Section 274 of the Atomic Energy Act of 1954. The paper will discuss past and current legislation, rulemaking, and regulatory developments pertaining to resolving the transuranic waste discrepancy and providing a pathway to disposition GTCC LLW. (authors)

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DOE OSTI2012

Generic Repository Concepts and Thermal Analysis for Advanced Fuel Cycles - 12477

A geologic disposal concept for spent nuclear fuel (SNF) or high-level waste (HLW) consists of three components: waste inventory, geologic setting, and concept of operations. A set of reference geologic disposal concepts has been developed by the U.S. Department of Energy (DOE), Used Fuel Disposition campaign. Reference concepts are identified for crystalline rock, clay/shale, bedded salt, and deep borehole (crystalline basement) geologic settings. These were analyzed for waste inventory cases representing a range of waste types that could be produced by advanced nuclear fuel cycles. Concepts of operation consisting of emplacement mode, repository layout, and engineered barrier descriptions, were selected based on international progress. All of these disposal concepts are enclosed emplacement modes, whereby waste packages are in direct contact with encapsulating engineered or natural materials. Enclosed modes have less capacity to dissipate heat than open modes such as that proposed for a repository at Yucca Mountain. Thermal analysis has identified important relationships between waste package size and capacity, and the duration of surface decay storage needed to meet temperature limits for different disposal concepts. For the crystalline rock and clay/shale repository concepts, a waste package surface temperature limit of 100 deg. C was assumed to prevent changes in clay-based buffer material or clay-rich host rock. Surface decay storage of 50 to 100 years is needed for disposal of high-burnup LWR SNF in 4-PWR packages, or disposal of HLW glass from reprocessing LWR uranium oxide (UOX) fuel. High-level waste (HLW) from reprocessing of metal fuel used in a fast reactor could be disposed after decay storage of 50 years or less. For disposal in salt the rock thermal conductivity is significantly greater, and higher temperatures (200 deg. C) can be tolerated at the waste package surface. Decay storage of 10 years or less is needed for high-burnup LWR SNF in 4-PWR packages, while 12-PWR packages could be emplaced after 40 years or less. HLW from reprocessing LWR UOX fuel or metal fuel from fast reactors, could be disposed of in salt after 10 to 50 years of decay storage depending on the specific composition and other factors. For the deep borehole disposal concept no near-field temperature limits are recognized because no performance credit is taken for waste form or waste package integrity, or containment by the near-field host rock. These results show the key differences in thermal management strategies available to the U.S. repository program, given the range of disposal concepts. A host medium such as salt with greater thermal conductivity and peak temperature tolerance could shorten decay storage by 50 years, or facilitate the use of larger waste packages. The LWR UOX SNF evaluated in this study represents that which could be produced in the coming decades. The existing, lower burnup used fuel that is presently in storage at many LWR locations across the U.S. is significantly cooler, and analyses of this type could be used to show that disposal is possible with less decay storage or larger waste packages. We note that while the temperature limits and waste package capacities used in this study are similar to those used internationally and in past U.S. studies, they might be increased as the result of ongoing research and development activities. This study selected enclosed emplacement modes to conform with disposal concepts developed internationally and previously in the U.S. Open modes (such as that proposed for a repository at Yucca Mountain) afford additional flexibility in waste management and the necessary investment, because the same facility serves both storage and disposal functions. Use of open modes, and combined analysis of storage, transportation, and disposal functions, are appropriate to consider in future studies of this type. (authors)

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DOE OSTI2020

Codisposal Waste Package Loading Options for DOE SNF and HLW PPT

The U.S. Department of Energy (DOE) is responsible for managing spent nuclear fuel (SNF) that is currently in, or will in the future come into, its possession. DOE must continue to safely store that SNF, transport it to an interim storage site or a repository, and dispose of it. These fuels come from a wide range of reactor types with various cladding materials and enrichments. Many of these reactors, now decommissioned, had unique design features, such as core configuration, fuel element and assembly geometry, moderator and coolant materials, operational characteristics, and neutron spatial and spectral properties, resulting in a large diversity of reactor and fuel designs. Because of the wide variety and conditions of SNF, a robust DOE Standard Canister was proposed that would confine radionuclides and preclude moderator. The DOE Standard Canister had four variations: 3.05-meter (10-foot) or 4.57-meter (15-foot) length, and 45.7-cm (18-inch) or 61.0-cm (24-inch) diameter. For ultimate disposal in the Yucca Mountain Repository, these canisters were to be grouped with 61.0-cm (24-inch) diameter high level waste (HLW) canisters in a 2.13-meter (84-inch) diameter co-disposal waste package. The smaller 45.7-cm (18-inch) diameter DOE Standard Canister could be placed in the middle of five HLW canisters. The larger 61.0-cm (24-inch) diameter DOE Standard Canister would take the place of one of the five HLW canisters on the outer ring in the co-disposal waste package. No DOE Standard Canisters have been loaded. A preliminary evaluation has estimated the number of elements of a fuel type that can fit into the different sizes of the DOE Standard Canister, but no definitive loading configuration has been selected. Changing the loading configuration could impact the number of loadable DOE Standard Canisters and the number of co-disposal waste packages needed for eventual disposition. This paper conveys the ranges of DOE Standard Canisters and HLW canisters that may be produced under certain conditions. It also examines the differences in the estimated number of co-disposal waste packages produced for eventual disposal when using different loading strategies in the DOE Standard Canister for Advanced Test Reactor (ATR), Peach Bottom, and High Flux Isotope Reactor (HFIR) SNF. Changing the loading configurations of ATR, Peach Bottom, and HFIR SNF slightly impacted the number of co-disposal waste packages that may be needed for ultimate disposal. The change in loading configuration was more impactful when a different canister was used, as opposed to varying the number of elements that could fit inside the same size canister. In one case, a reduction of co-disposal waste packages could be achieved by allowing mixing of short HLW canisters with long DOE Standard Canisters. The main conclusion from this analysis is that the ratio between HLW canisters and DOE Standard Canisters will drive the total number of co-disposal waste packages. If too many HLW canisters (i.e., more than five times the number of 18-inch DOE standard canisters) or DOE Standard Canisters are produced, some co-disposal waste packages may not have all positions filled. A co-disposal waste package may be filled with all HLW with no DOE Standard Canister, or a co-disposal waste package could be filled with a single DOE Standard Canister. A ratio that does not closely align to optimum could allow for the design of waste packages that hold just HLW canisters or just DOE SNF canisters.

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DOE OSTI2019

International Low Level Waste Management Practices and Technical Bases for Very Low Level Waste Classification - 19415

Different countries around the globe use varying schemes for the classification of radioactive waste. Many countries use adaptations of the International Atomic Energy Agency (IAEA) waste classification scheme. The United States uses the Class A, B, and C waste classification scheme that is loosely based on the IAEA concepts. Classification may be based on concentration limits or dose rate limits or a combination of concentration and dose rate limits. In some countries the existence of very low-level waste categories allows nuclear power plant waste to be disposed of in hazardous waste or shallow land burial sites separate from low and intermediate level waste disposal facilities. In the United States, the very low-level waste (VLLW) classification does not exist. Nuclear power plants use alternative disposal procedures under 10CFR20.2002 and other licensed processes in agreement states in order to dispose of very low activity waste in hazardous waste disposal sites. [1] Disposal of VLLW in hazardous waste or separate VLLW facilities is generally more cost effective than disposing of waste in radioactive waste sites and is protective of the health and safety of the public. This can become an important factor when large volumes of very low activity waste are generated during decommissioning of nuclear power plants. The Electric Power Research Institute (EPRI) Radiation Safety (RS) Program has developed technical bases for the establishment of a very low-level waste category. EPRI studied how VLLW waste is defined, managed, and disposed of in various countries. This included the study of disposal site design, characteristics, and how these sites protect the health and safety of the environment and the public from low levels of radioactivity. EPRI conducted a cost benefit analysis to understand the economic impact of disposing VLLW in hazardous waste disposal sites versus low-level waste disposal sites. Finally, EPRI calculated potential radionuclide limits that may be used to define very low-level waste. The purpose of this work was to provide an example of how one might approach defining limits for VLLW. The same or similar approaches could be applied to derive radionuclide limits for different dose limit objectives. This paper presents the results of research that explored the waste management and disposal practices at six countries around the world with additional focus and details on the implementation of the very low-level waste category. (authors)

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DOE OSTI2017

Conceptual Design for a Short-Lived Low- and Intermediate-Level Waste Repository at the Al Tuwaitha Nuclear Center, Baghdad, Iraq - 17189

The European Commission under their Instrument for Nuclear Safety Cooperation (INSC) has initiated a project supporting Iraq's efforts for the safe disposal of short lived low- and intermediate-level radioactive waste (SL-LILW) arising from the decommissioning of Iraqi nuclear installations. The INSC project includes both the development of an initial conceptual level basic design and preliminary safety assessment (PSAR) report and the subsequent detailed engineering design and supporting intermediate safety assessment report (ISAR). The project is being executed by a Consortium led by NUKEM Technologies Engineering Services GmbH and joined by DBE Technology GmbH of Germany; JAVYS a.s. of Slovakia and MCM Environmental Services Ltd of the United Kingdom. The current paper focuses on the development of the conceptual level basic design. In previous decades Iraq had a significant nuclear program which encompassed ten nuclear sites spread out across the country. The largest site was located at the Al-Tuwaitha site near Baghdad where 18 nuclear facilities were located. The facilities suffered significant damage from bombing and subsequent looting during the first Gulf War in 1991. Currently, the Ministry of Science and Technology (MoST) is the radioactive waste management operator in Iraq and also the license holder for the ten Iraqi nuclear sites, all of them heavily damaged. Because of the ongoing security concerns in Iraq MoST has selected the Al Tuwaitha Nuclear Center as the preferred and best suited location for development of an SL-LILW repository. As an initial step in development of the conceptual design various repository design options for the site were examined and assessed. The assessment considered similar facilities that have been implemented internationally. At the conclusion of the assessment and in conjunction with the Iraqi Beneficiaries from MoST, a near-surface vault design was selected for the facility. The design of the facility relies on experience gained from existing low-level waste repositories at Centre de l'Aube in France and the El Cabril solid radioactive waste disposal facility in Spain as well as on more recent experience gained in the design of the Bulgaria National Disposal Facility (NDF) under construction near the Kozloduy Nuclear Power Plant. Similar to these reference designs, the concept for the Al Tuwaitha Disposal Facility (ATDF) will rely on a series of reinforced concrete vaults into which containerised waste will be placed. Each vault will consist of two chambers designed specifically to hold the waste. Central drainage points in each waste chamber will lead to an infiltration water collection system designed to detect any failures in the waste containment. Each disposal vault will be sealed after filling by installation of a reinforced concrete closure slab. Once all of the waste has been disposed of the site will be decommissioned and a multilayer protective cover will be installed over the sealed disposal vaults. The multilayer cover will act to further isolate the waste from the environment by limiting the access of water to the waste. Due in part to the presence of shallow groundwater at the site and to avoid potential damage to the facility by flooding from the nearby Tigris River the concept for the Al Tuwaitha Disposal Facility requires construction of the disposal vaults on a raised platform several meters above ground level. Construction of this platform will be integrated into the conceptual safety design and materials will be selected which further enhance the isolation function of the disposal system by hindering or delaying the release of radionuclides to the environment. The infiltration water collection system, which will be constructed in a specifically designed accessible gallery embedded and fully integrated into the platform, will provide access to inspection points located beneath each disposal chamber. The ATDF when completed will be a further demonstration of the adaptability of the disposal vault concept for use in isolating SL-LILW from the environment. (authors)

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DOE OSTI2017

Streamlining and Transparently Automating Complex Waste Determinations at Los Alamos National Laboratory - 17286

Environmental programs at Los Alamos National Laboratory (LANL) generate a large amount of sampling activities that can result in waste that needs to be characterized and properly disposed. Since 2011, sampling results are managed in the cloud-based Environmental Information Management (EIM) database, which contains over 16 million current and historical data records. In many cases, the database's analytical records are used to characterize and dispose waste from environmental operations. Three factors contribute to streamlining of the waste disposition operations: (1) if possible, use of the existing analytical data for waste characterization to avoid additional laboratory analysis; (2) reduction in time spent on data evaluation for various waste acceptance criteria; and (3) choice of the proper and least expensive disposal alternative. Use of the existing analytical data, when allowed by law, creates some paradoxical problems: when chemical analysis is requested with the purpose of characterizing waste, it creates only 100 to 400 analytical records. However, use of exiting analytical data at LANL may lead to evaluation of tens of thousands of analytical records, with many records for each single analyte. This occurs because data were collected for environmental characterization and not waste characterization. To streamline data analysis for waste disposition using existing data, LANL has created a Waste Data Summary (WDS) module in the site's EIM database. Recognizing that waste determination is performed by groups outside the environmental program at LANL, the WDS allows for logically guided data selection for a specific waste determination (solids, liquids, radiological, etc.) and reduces potentially thousands of rows of data into simple formatted data reports that are easy to comprehend and process. This allows the waste evaluators to review data using multiple approaches to make their own determination on the most appropriate disposal method. There are four basic principles of WDS operation: (1) The highest detected analytical result is evaluated for the selected samples (or the lowest undetected result if there are no detected results); (2) All approved disposal pathways are evaluated; (3) Any failures of the Waste Acceptance Criteria are visibly highlighted to focus the attention of the evaluator; and (4) The decision of the disposal path is left to the trained waste evaluator (WDS does not choose or suggest any disposal paths). The summary results are presented as the set of short data reports. These reports can be divided into two groups: general comparison reports and disposal pathway specific forms. The general comparison forms show all Resource Conservation and Recovery Act (RCRA) analytes with pass/fail evaluation and with preference for toxicity characteristic leaching procedure (TCLP) results (or 20x rule) for solid samples, the potential RCRA hazard codes assignment for detected analytes, and all rad results. The detected chemical and radioactivity results are compared to the background values and to Soil Screening and Soil Action limits (SSL/SAL limits). Disposal pathway specific forms follow Waste Acceptance Criteria requirements for all disposal pathways available at LANL. The programming of the WDS tries to maximize existing data by analyte substitutions (like using gross alpha for necessary alpha emitting isotopes) or by making calculations when analytical results do not agree with legal requirements (like total chlordane in the RCRA regulations, which is most often analyzed and reported as individual isomers). All forms may be annotated by the waste evaluator to provide additional information to support decision making. Upon final decision of a disposal path, the complete dataset used in the evaluation and the WDS forms with annotations is 'frozen' as an un-editable PDF export to document the data used in the waste disposal determination. The streamlined process results in less work for waste evaluators and reduces overall labor and analytical costs. By comparison, a full review of a 10,000 record parent dataset may take a waste evaluator over a week to review, whereas with the WDS module, the task can be accomplished in less than one hour with clear, reproducible results. These results can be reviewed and checked at any time by quality assurance personnel, thus ensuring transparency in the decision making process. LANL believes this approach has broad applicability to other sites where waste determination is based on review of existing data collected for more generalized environmental purposes. (authors)

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