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DOE OSTI2017

Radioactive Waste Management in a Treatability and Optimisation Perspective - 17527

In order to meet recycling criteria or to fulfil Waste Acceptance Criteria (WAC) for disposal there are many different management concepts for radioactive waste that can be applied. Which concept or set of concepts to use to optimize waste management depends on several factors, among them, the regulatory framework for recycling, the availability of a disposal facility and the acceptance criteria for disposal. The required pre-treatment and other pre-disposal activities should be carefully analyzed for each of the disposal options, as they are significant contributors to the total cost for managing the waste. Waste Acceptance Criteria are living documents that should be updated when regulatory framework changes as well as when new treatment methods and new packaging and barrier systems develop. Each waste treatment facility, has its own specific limitations and therefore its own WAC/CFA (conditions for acceptance), which depends both on the plant design and national legislation. Many waste owners, countries or licensees, have their preferred, and in some cases well established, waste routes for certain waste streams. In some countries, the full chain from waste treatment through to disposal of radioactive waste is in place, whilst other countries still are in the process of establishing their repository program. The worst case scenario for any licensee or waste owner is waste which is not possible to qualify for disposal. This can be due to poor or degraded waste containers, unspecified content in waste packages, or wastes that are not suitable for disposal in their current form. Such deficiencies result in expensive actions related to waste retrieval, new characterisation and/or costly investigations. The EDF Group has a fleet of waste treatment facilities and has, for decades, provided waste treatment services to domestic and international customers. This paper will discuss important waste owner considerations and EDF experiences related to the optimization of the waste management process. This will be from both a treatment and an end-state perspective. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

From Dealing with Legacy Waste to Avoiding It - 17432

When dealing with radioactive waste management and its regulation, there are various national approaches. At the same time there is a growing tendency towards consideration of fundamental principles and standards thus forming the underlying basics in detailing the customized regulatory framework for specific applications. Dealing with legacy waste poses some particular problems as at the time of its production current safety principles were sometimes not well established. Considering today's standards this frequently requires remedial actions. Usually, this involves considerable effort and cost as well as radiation exposure for the involved personnel. Resolving legacy waste related problems not only benefits from an established regulatory framework but allows conclusions on how deal with legacy waste problems in general. The most important lessons learned from such operations might be how to prevent the production of legacy waste in the first place, thus avoiding further problems. By reflecting on practical applications from a number of countries the presentation is set to carve out the crucial recommendations from internationally recognized standards that give a substantial dividend in terms of not only time and budget but most of all safety and radiation exposure when properly implemented. The scope of the examples given ranges from dealing with legacy waste to avoiding it and includes recommendations towards essential features of best practice as seen from the perspective of an independent expert organization. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

CCUS Perspectives: Assessing Historical Contexts, Current Realities, and Future Prospects

CCUS technologies are crucial solutions for mitigating climate change by reducing CO2 emissions from industrial operations and energy sectors. This review critically examines the current state of CCUS technologies, and highlights advancements, challenges, regulatory frameworks, and future directions. It comprehensively analyzes carbon capture methods, such as pre-combustion, post-combustion, and oxy-fuel combustion capture, while comparing their efficiencies and limitations. The review also explores carbon utilization techniques, such as direct and indirect utilization, emphasizing their potential applications and technological constraints. Additionally, it assesses various carbon storage methods, focusing on geological, ocean, and mineralization storage, and discusses their capacity, feasibility, and environmental implications. The study reviews the policy and regulatory frameworks, economic viability, market trends, and environmental sustainability of CCUS. By identifying research gaps and recommending future research priorities, this review aims to guide the development of more efficient/effective, and cost-effective CCUS technology, ensuring their role in a sustaining low-carbon future. This review provides a forward-looking perspective, a critical and interdisciplinary analysis that assesses the current state of CCUS technologies, and further provides a roadmap for future development.

Energy & Fuels↗
DOE OSTI2024

Soil Desiccation Treatability Testing at BC Waste Disposal Cribs

During Hanford’s production period, low-level waste products generated from chemical processing of uranium fuel rods were discharged directly to the ground through a system of cribs and trenches located in the 200-BC-1 Operable Unit (OU). The site consists of 6 cribs and 20 trenches that received more than 117,000 m3 of radioactive liquid waste discharged to the soil. These unlined infiltration galleries held volumes of liquid waste while it seeped into the ground, with the understanding that the 100 m (330 ft) thick vadose zone in the area would effectively capture the effluent and prevent groundwater impacts. A conceptual model showing the operation of cribs and trenches is shown in Figure 1. Data show effluent from the 26 cribs and trenches containing about 410 curies of Technetium-99 (Tc-99) is primarily located between 30 m and 70 m (98 ft and 230 ft) depth (Corbin et al., 2005; Ward et al., 2004). Despite no evidence indicating that the contamination has reached the groundwater at BC cribs and trenches, the mobility of Tc-99 had been demonstrated in laboratory tests and was recognized as a threat to groundwater at the site. Using data from numerical models, laboratory analyses, field investigations, and information on historical discharges, the EPA and Ecology identified Tc-99 and U contamination of the vadose zone as a remediation priority. The U.S. DOE was notified by EPA and Ecology regarding risks associated with Tc-99 contamination in a letter requesting development of a strategy for improved methods to understand the nature and extent of vadose zone contamination, specifically Tc-99, and to develop remedial options for addressing such contamination. To develop the appropriate technology for characterizing, remediating, and monitoring the deep vadose zone Tc-99 contamination, the U.S. DOE worked with the EPA and Ecology to create a Treatability Test Plan under a Remedial Investigation/Feasibility Study (RI/FS) for the Hanford 200 Areas. Under this RI/FS, it was determined that a treatability test for soil desiccation should be carried out as it was identified as a promising in-situ treatment technology for mitigating risks posed by Tc-99 contamination to the groundwater table. The BC Cribs and Trenches site was identified as a representative site for Tc-99 and U contamination and selected for the soil desiccation treatability test. In this chapter, we summarize the overlying regulatory framework of RI/FS and treatability tests and illustrate how development and experimentation supported the evaluation of selected remedies. We briefly discuss the RI/FS for the 200 Areas of the Hanford Site and focus on the soil desiccation treatability testing performed at the BC cribs and trenches site under the Deep Vadose Zone Treatability Test Plan for the Hanford Central Plateau (DVZ-TT). The DVZ-TT is one component of the remedial investigation/feasibility study for the Hanford 200 Areas and represents the underlying regulatory framework that drives site operations towards records of decision and site closure.

Mangel, Adam R.↗
DOE OSTI2019

The Challenges of Repository Operation in an Existing Mine - 19339

A generally accepted principle related to long term management of radioactive waste/ spent fuel is the isolation in a geological repository. In Germany, the disposal operation of low-level waste (LLW) and intermediate-level waste (ILW) started in the Asse II mine in 1967 and in the former German Democratic Republic in the Morsleben repository in 1971. After the German reunification, the responsibility for the existing sites changed several times until the Bundesgesellschaft fuer Endlagerung (BGE) became new license holder for both sites in 2017. The BGE is responsible to carry out the operative tasks of searching for a high-level waste (HLW) repository, the planning and construction of the Konrad repository for the emplacement of low- and intermediate-level waste (LILW), the operation of the Asse II mine including the retrieval of the emplaced waste, and the preparation for the closure of the Morsleben repository. Nevertheless, the challenges of repository operation in an existing mine have not changed in all these years, as the regulatory framework is mainly written for Nuclear Power Plants (NPPs) and is difficult to adapt for a repository mine. In particular, conditions in a mine are substantially different from 'clean' above-ground nuclear installations. Dust due to mining activities, relatively high temperatures, and a corrosive atmosphere in a salt mine might be present. Therefore, the usability of the repository for several decades during the different operational phases, i.e. planning, construction, emplacement, and closure, can only be ensured by continuous maintenance. Furthermore, the atomic law is applicable for above-ground facilities and the mining law for mines of every description. The difficulty of geological repositories is to ensure that both laws are going to be applied above and below ground with no regulatory framework tailored especially for such sites. A special case is the Asse II mine where the mining law was applied from the very beginning, before the atomic law needed to be established later on and now requires the retrieval of the radioactive waste before decommissioning. In this presentation, similarities and differences in the legal framework of the repositories in the responsibility of BGE and some operational experiences throughout the years are described. This applies to the radiation protection regime, the excavation ratio, the host rock stability and the backfilling concepts of the mining areas. Further challenges arise from the demonstration of long term safety according to the state-of-the-art of science and technology for either existing repositories/ mines or those under construction or planned. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Evaluating Probability of Containment Effectiveness at a GCS Sites using integrated assessment modeling approach with Bayesian decision Networks

Improved scientific and engineering understanding of the behavior of geologic CO2 storage together with established regulatory framework and incentive structures raise the prospects for accelerated, large-scale deployment of this greenhouse gas emissions reduction approach. Incentive structures call for the establishment of appropriate verification and accounting approaches to support claims of the integrity of a geologic storage complex and to justify taking credit for long-term storage. In this study, we present a framework for assessing the probability of containment effectiveness over the lifetime of a geologic carbon storage site (e.g., after 70 years of injection and post-injection site performance) using forward stochastic model realizations based on site characterization data and using a monitoring-informed Bayesian network based on hypothetical detectability from surface seismic surveys over the site injection and post-injection phases. The National Risk Assessment Partnership’s open-source Integrated Assessment Model (NRAP-Open-IAM) was utilized to develop an ensemble of 10,000 a priori stochastic forecasts of CO2 containment. Those simulations were used to train the Bayesian network model to estimate the prior probabilities of the CO2 leakage mass into overlying, monitorable aquifers considering the uncertainties in the reservoir properties, permeability of potentially leaky wells and the overlying aquifers. The conditional probabilities in the Bayesian network were either learned from the NRAP-Open-IAM simulations or derived from the predefined detection thresholds for the monitoring method. Observations obtained from monitoring, over time during the site operation phases were then used to generate updated posterior probabilities of containment (and any loss from containment) in the Bayesian network by propagating the prior probabilities through the conditional probabilities. We demonstrate how to construct and use the Bayesian network for verifying the long-term storage complex effectiveness informed by monitoring based on the NRAP-Open-IAM simulations previously developed for the FutureGen 2.0 site. This approach may have relevance for stake holders to demonstrate secure geologic storage, provide a defensible, probabilistic approach to claim credit for geologic storage, and to estimate the likelihood that any fraction of the claimed credit may need to be refunded to the creditor based on available monitoring information.

Bayesian network, Risk assessment, Monitoring, car↗
DOE OSTI2017

Post-Fire Safe Shutdown Capability. An Inspector's Handbook. Draft Report

The Nuclear Regulatory Commission's (NRC) current fire protection regulatory framework was initiated in response to a cable fire that occurred at the Browns Ferry Nuclear Power Station in 1975. Following the issuance of several guidance documents, in 1981 the Commission codified deterministic requirements to provide reasonable assurance that fire would not jeopardize reactor safety (10 CFR 50.48 "Fire Protection" and Appendix R to 10 CFR 50). Since then, the results of plant operating experience, NRC inspection activities and cable fire testing programs have served to clarify the application of these requirements. In 2004, the NRC amended 10 CFR Part 50.48 to add a new subsection, 10 CFR 50.48(c), that endorses, with exceptions, the National Fire Protection Association's 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants — 2001 Edition," as a voluntary alternative for demonstrating compliance with Appendix R. To date, approximately half of all operating plants have transitioned to a risk-informed approach under 10 CFR 50.48(c). Regardless of the selected approach, each operating nuclear power plant should establish a comprehensive fire protection program that includes features needed to minimize the likelihood and consequence of firesl. To achieve this objective, fire protection programs integrate the NRC's long-held safety concept of defense-in-depth (DID) by providing multiple barriers against potentially unacceptable consequences of fire. To minimize the frequency and size of fires, the first lines of defense include controls for fire prevention, such as controlling combustibles and ignition sources and fire mitigation features, such as fire detection and suppression systems. In the unlikely event that a fire should start and continue to grow in spite of these features, appropriate fire protection features, such as fire-rated barriers are provided as necessary to ensure the accomplishment of essential shutdown functions. This handbook was specifically developed for NRC inspectors that may be assigned to perform elements of Inspection Procedure (IP) 71111.05T, "Fire Protection (Triennial)," including, (a) "Protection of Safe Shutdown Capabilities," (e) "Alternative Shutdown Capability," and (f) "Circuit Analyses." Thus, the document concentrates solely on the final element of defense-in-depth— ensuring reactor safety will not be compromised in the event of a serious fire. In addition to describing how the NRC's regulatory framework has evolved since the Browns Ferry Fire, the handbook illustrates the content and scope of various analyses typically referenced in a plant's fire protection licensing basis and includes techniques and insights for assessing the plant's conformance to those criteria.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Integrated Commercial Carbon Capture and Storage (CCS) Prefeasibility Study at Rock Springs Uplift, Wyoming (Final Report)

To establish the potential of safe, long-term commercial-scale carbon capture & storage (CCS) in southwestern Wyoming, the University of Wyoming formed a CCS coordination team (CCT) to complete a Phase I CarbonSAFE pre-feasibility study adjacent to the region’s largest coal-fired power plant, the Jim Bridger Plant (JBP). The study’s primary objective was to evaluate the potential of storing 50+ million metric tons of anthropogenic carbon dioxide (CO 2 ) in stacked reservoirs over a period of 25 years. In addition, this integrated study evaluated related aspects of carbon capture utilization & storage (CCUS) at the study site, including: (1) the CO 2 character of post-combustion flue gas from JBP and the technical capacity for retrofitting proven commercial-scale CO 2 capture technology relative to CO 2 source character and technological constraints; (2) utilization of the existing CO 2 pipeline network in the immediate vicinity of JBP for enhanced oil recovery (CO 2 -EOR) and utilizing that network as part of a regional hub for CO 2 storage from other anthropogenic sources; (3) assessment of the challenges and benefits of meeting all of CarbonSAFE programmatic Phase I goals relative to Wyoming’s carbon management regulatory framework; and (4) assessment of the community and environmental factors that may impact CCS at JBP. With respect to storage, we analyzed the potential of stacked Mesozoic reservoirs at the study site, focusing on the Entrada and Nugget sandstones. The reservoir response to injection simulations varied greatly between formations, though they both were capable of CO 2 storage. The Entrada Sandstone is relative thin (~40’) and heterogeneous; storage using one injection and production well was limited to 1.2 Mt over 25 years at the study site. The Nugget Sandstone is much thicker (> 400’) and generally homogeneous; storage using one injection and production well approached 15.0 Mt over 25 years; this scenario suggests that the storage potential in the Nugget Sandstone is close to the estimated P90 calculation of 9.6 Mt/mi 2 as models indicate a storage potential of 8.3 Mt/mi 2 at saturation. Both reservoirs’ storage capacity benefited greatly from coupling an injection well with an offset production well; total storage capacity increased and reservoir pressure stayed below critical thresholds. This indicates that implementing a pressure management strategy at the study site would optimize storage and decrease risk. Seals associated with the targeted reservoirs were shown to be capable of retaining the potential storage volumes, and the site benefits from multiple seal redundancies (i.e. > 7,000’ of seal formations). Furthermore, assessment of different reservoir variables (e.g., fluid composition, pressure, etc.) indicates all reservoirs at the study site are isolated. Storage assessments from Mesozoic reservoirs were coupled with simulations from the site’s Paleozoic reservoirs (the Weber Sandstone and Madison Limestone), which were the focus of previous study. The fully integrated stacked reservoir pre-feasibility studies suggest that the Nugget and Madison formations alone could safely store 50 million metric tons of CO 2 within the study site. Furthermore, both formations are true saline reservoirs near JBP, making them ideal target reservoirs for CO 2 sequestration. Analysis of Wyoming’s laws and regulations confirm that the State has a favorable policy environment for CCS/CCUS. The project’s economic models estimated that the following revenues collectively are sufficient to finance the project over its lifespan if CO 2 utilization is coupled with CCS via the following strategies: (1) sales of CO 2 for EOR (approximately $69 million/year); (2) use of CO 2 tax credits such as amended §45Q and §48A (approximately $484 million total); and (3) sales of low-carbon electricity and marketable carbon offset/credits into carbon-constrained markets (approximately $11-$17 million/year) to West Coast consumers. These revenue estimates are broadly consistent with other studies that concluded arbitrage in JBP retail electricity sales between California and Wyoming could help to support the cost of deploying CCS at JBP. The models also indicate that maintaining a successful business case for long-term CCS -- wherein associated capture, transportation and operational costs are fully addressed at JBP -- is reliant on numerous external factors that complicate predictions. This study suggests that this site meets the CarbonSAFE program’s requirements of being able to feasibly store 50+ million metric tons of CO 2 over 25 years within the site’s stacked reservoirs, especially if coupled with pressure management strategies. This study suggests that the Nugget Sandstone and the Madison Limestone are two of the best reservoirs in the State with respect to overall storage capacity. In addition, the site’s proximity to existing CO 2 transportation networks, CO 2 -EOR opportunities, ability to sell to markets that value low-carbon electricity, Wyoming’s existing carbon regulatory framework, and a public that is well-educated with respect to energy markets and issues collectively increase the potential for implementing commercial-scale CCS adjacent to the JBP. With respect to the available technical and non-technical site data, and the reservoir quality at the study site, we suggest that this site is capable of moving directly into Phase III of the CarbonSAFE program.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2019

Communicating Technical Information: Different Approaches for Different Audiences - 19399

The Canadian Nuclear Safety Commission (CNSC), Canada's independent nuclear regulator, has a 70- year history of regulating nuclear safety in Canada. Part of the CNSC's mandate is to disseminate objective scientific, technical and regulatory information to the public. The CNSC prides itself on being transparent with the public, and openly sharing scientific, technical and regulatory information on its regulatory activities with stakeholders. We value the public's feedback, and have a robust public participation program to allow for public involvement in the CNSC's activities. However, different audiences have different needs based on their level of technical knowledge and expertise. The CNSC's experience with stakeholder interactions has demonstrated the need to balance the dissemination of technical information through different communication platforms, in order to satisfy the needs of various intended audiences. This paper will focus on the radioactive waste management communications methods that the CNSC has implemented for two distinct audiences: 1. Audiences with little or no technical background: Communications with this audience requires techniques to make highly technical concepts more understandable. The focus is on educating about matters related to radioactive waste management, and informing about the safety aspects behind the regulation of radioactive waste. 2. Audiences with previous knowledge and experience with nuclear matters: Communications with this audience requires techniques to ensure that as much technical information and data as possible is available to inform stakeholders. The focus is on giving access to studies, research and raw data, in order to allow this audience to inform themselves, and potentially bring new perspectives to the CNSC's attention that could be used in regulatory oversight activities. In its capacity to disseminate scientific and technical information to the public as required by legislation, the CNSC holds a unique position compared with other Canadian regulatory bodies. The CNSC is one of the only Canadian regulators which: - Has a public hearing process which enables the public and Indigenous groups to express their issues and concerns directly to the decision-maker (the Commission) - Publishes regular regulatory oversight reports (report card on industry) that assess the safety performance of all its licensees - Publishes responses to all public comments received during consultation periods in our environmental assessment processes as well as our regulatory framework processes, including draft regulatory documents, discussions papers and proposed regulations - Has an independent environmental monitoring program, allowing the public access to environmental data. The program also seeks public and Indigenous input with respect to proposed environmental sampling campaigns, including suggestions on the samples to be taken and sampling locations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Applicability of the Milestones Approach to Deployments of Transportable Nuclear Power Plants (TNPPs)

Transportable nuclear power plants (TNPPs) can provide potential benefits to countries embarking on nuclear programs, offering reduced infrastructure requirements, shorter timeframes for implementation, cost savings and greater deployment flexibility than larger conventional reactors. However, the deployment of a TNPP in a Host State comes with the obligation to establish sufficient regulatory, institutional, and technical infrastructure, which, among others, includes a legal and regulatory framework and a competent regulatory body to implement a State’s safeguards obligations. This paper considers how the unique technical and deployment features of TNPPs may affect the process of preparing for and implementing safeguards in nuclear newcomer countries. Evaluating this issue through the lens of the IAEA’s Milestones Approach, this paper discusses some potential implications arising from the shortening of some milestones phases due to reduced construction or licensing time for TNPPs, and the need for increased cooperation between Host States and Supplier States in preparing for and meeting certain safeguards obligations. These considerations are potentially relevant to various stakeholders: newcomer States considering TNPP deployment; the States and companies that supply such reactors; as well as organizations that support international safeguards capacity building.

Siserman-Gray, Ioana-Cristina↗
DOE OSTI2018

U.S. Nuclear Regulatory Commission Historical Perspectives of Very Low-Level Radioactive Waste - 18367

The United States Nuclear Regulatory Commission (NRC), in Title 10 of the Code of Federal Regulations Part 61 (10 CFR Part 61), 'Licensing Requirements for Land Disposal of Radioactive Waste', provides licensing procedures, performance objectives, and technical requirements for the issuance of licenses for the land disposal of low-level radioactive waste. Low-level radioactive waste is divided into four categories based on its potential risk and need for isolation. Class A is the lowest category of low-level radioactive waste and poses the least risk. A large percentage of Class A low-level waste contains very low radionuclide concentrations and does not pose a significant health and safety risk. In the current regulatory framework, this waste, referred to as low-activity or very low-level waste, is managed in the same manner as much more hazardous waste. In 1986 and 1990, as a result of the Low-Level Radioactive Waste Policy Amendments Act of 1985, the NRC issued the Below Regulatory Concern (BRC) policy statement. This policy statement established an implementation plan regarding the expeditious handling of rulemaking petitions 'to exempt specific radioactive waste streams from regulation due to the presence of radionuclides in sufficiently low concentrations or quantities as to be below regulatory concern.' The BRC policy statement provided a consistent risk framework when considering regulatory exemption decisions, such as: (1) the release for unrestricted public use of lands and structures containing residual radioactivity; (2) the distribution of consumer products containing small amounts of radioactive material; (3) the disposal of very low-level radioactive waste at other than licensed disposal sites; and (4) the recycling of slightly contaminated equipment and materials. Due to concerns with the BRC policy that were raised by Congress, the States, the public, and others, Congress enacted the Energy Policy Act of 1992, which revoked the NRC's 1986 and 1990 BRC policy statements and led to the NRC's formal withdrawal of the BRC policy in 1993. Stakeholders raised concerns on NRC's potential deregulation of radioactive material. Additionally, in March 2005, the NRC staff proposed a rule to amend 10 CFR Part 20, 'Standards for Protection Against Radiation,' to include radiological criteria for controlling the disposition of solid materials with very small amounts of residual radioactivity. The NRC staff also proposed a limited disposition approach for the disposition of solid material, limiting the release of solid material meeting a 0.01 milli-Sievert per year (mSv/yr) [1 milli-rem (mrem/yr)] dose criterion, from licensed control. In June 2005, the Commission disapproved publication of the proposed rule and deferred the rulemaking until a later time, stating that the NRC's existing approach of reviewing specific cases on an individual basis was fully protective of public health and safety. This rulemaking effort was discontinued by the Commission in 2016. Recently, several nuclear power plant licensees have announced early decommissioning of their facilities. Some stakeholders have expressed concerns about the capacity and disposal costs associated with using the four existing low-level waste disposal sites for very low levels of radioactive waste (i.e., waste at the low end of the Class A category). A more risk-informed, performance based approach to waste disposal would allow more flexibility in managing this waste commensurate with its risk and hazard. A new approach would also allow the NRC to align with international standards for this type of waste. The NRC is presently conducting a scoping study to identify the actions the agency should take to strengthen its regulatory framework for very low-level waste disposal. The Very Low-Level Waste (VLLW) scoping study will consider divergent stakeholder comments; lessons learned from the BRC policy statements and the proposed rule on radiological criteria for controlling the disposition of solid materials; and best practices of other countries with respect to VLLW disposal. The NRC will also engage other government agencies to discuss regulatory options for VLLW. This paper offers a historical overview of past NRC initiatives to manage the lower end of Class A waste in a manner that is risk informed and performance based. It examines previous endeavors and discusses the regulatory challenges that arose. It also addresses lessons learned and introduces NRC's current effort regarding very low-level radioactive waste disposal. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Reliability and Integrity Management Scoping Study

The U.S. Nuclear Regulatory Commission (NRC) is developing the regulatory framework and technical expertise to support regulatory review of advanced non-light water reactor (ANLWR) designs. The NRC staff expects most of these designs to use the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) Section XI, Division 2 “Requirements for Reliability and Integrity Management (RIM) Programs for Nuclear Power Plants” for developing and implementing pre-service inspection (PSI) and in-service inspection (ISI) programs. The NRC recently endorsed ASME BPVC Section XI, Division 2 (BPV XI-2) in Regulatory Guide (RG) 1.246. This new ASME code is yet to be used in any applications submitted for NRC review. This report provides an overview of the current state of knowledge and practices within the industry for the use of BPV XI-2 for the development and implementation of a PSI and ISI program for non-light water reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Application of Electrochemical Methods to Molten Salt Reactors: Draft TLR Documenting Assessment of Electrochemical Monitoring

Nuclear Regulatory Commission (NRC) is developing the regulatory framework and technical expertise to support regulatory review of advanced non-water reactors, including molten salt reactors (MSRs). In an MSR, it is essential that the salt chemistry be maintained in a desired range in terms of redox potential for reliable operation and mitigation of corrosion to structural materials in the reactor, particularly the reactor vessel and heater exchanger. Measuring the redox potential of the salt in the reactor would also allow for the material lifetimes to be predicted more accurately, and chemical issues to be diagnosed more quickly. In addition to the chemical composition analysis by ICP-MS, electrochemical methods including potentiometry and linear wave scanning (LSC) were also used in molten salt reactor experiment (MSRE) for redox potential monitoring purposes. Electrochemcial methods offers unique advantages such as quick turnaround in results and unique capability of in-line monitoring of redox potential, and are considered popular electroanalytical techniques that can be used to monitor redox potentials and salt chemistry including impurities. The last two decades have seen significant advances in science and engineering of electrochemical methods for potential application to molten salts. The primary goal of this report is to assist NRC in understanding the monitoring of the salt chemistry by electrochemical methods and provide NRC reviewers with necessary information and tools to support regulatory review of MSR designs. This reports consists of two majors parts—Part 1, chemical potential of molten salts and effects by fission process in MSR; Part 2, assessment of electrochemical methods for application to MSRs. The TRLs of the typical relevant electrochemical methods for molten salts were evaluated based on the DOE TRL guidelines and upon a review of the current status of the electrochemical methods for two typical salt systems, fluoride and chloride, for MSRs.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

EDF industrial strategy for the management of radioactive waste - 15204

In France, EDF is in charge of electronuclear power operation since the sixties and consequently has to manage radioactive waste. Radioactive waste management is based on the control of each industrial step, taking into account the classification by activity level and radioactive half life. The main steps are to limit waste quantities at the generation stage, sort out non-rad and recoverable items from the waste, recycle, treat and package the radioactive waste as soon as it is generated, limit the use of interim storage, prepare and achieve as soon as reasonably practicable conditioning and disposal. EDF aims to empower operators in charge of NPP operating and of decommissioning and to centralize interfaces with the stakeholders in order to pool management routes and to guarantee the consistency of our industrial policy. The French regulatory framework for the management of radioactive waste and material is set by the French 2006 planning Act, in line with the European Directive. The legal framework structures interactions between EDF and the regulatory bodies (Nuclear Safety Authority, General Directorate for Energy and Climate), the French National Radioactive Waste Management Agency (ANDRA) and other waste producer (mainly AREVA Company and CEA). Regarding short lived waste, management routes are in current operation for the large majority of this waste generated both by power plants in service and during decommissioning. EDF is working on new management route for metallic materials knowing the future evolution of deposits is the field of large refit and decommissioning activities. Regarding long lived waste, disposal development represents a major stake for EDF. EDF contributes to design studies of the geological disposal center and collaborates with ANDRA for the acceptance of future national waste disposal. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Regulatory Considerations Regarding Potential High Temperature Fluid Releases in Advanced Reactor Designs

The current effort is supported by the U.S. Department of Energy (DOE), Advanced Reactor Demonstration Program (ARDP) Regulatory Development, Regulatory Framework Modernization area, which seeks to address potential regulatory challenges for advanced reactor vendors that are currently or will soon be initiating the licensing process. In pursuit of this goal, this effort seeks to aid the advanced reactor industry and regulatory bodies in understanding and addressing the potential occurrence of high-temperature fluid releases in advanced reactor designs as part of licensing and regulatory oversight of operation. Improving the awareness and understanding of the behavior and potential consequences associated with high-temperature fluid release events can ensure that they are appropriately considered and addressed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Applying Technology and Experience to Ensure Safe In Situ Decommissioning - 19587

Permanent entombment of a radioactively contaminated facility as a decommissioning option (In Situ Decommissioning, ISD) has been completed for DOE facilities at the Savannah River Site (P and Reactors, HWCTR reactor), Hanford (U Canyon) and Idaho (CP-601/604),and is currently being considered for implementation at a limited number of selected DOE facilities as well as for the NPD and WR-1 reactors in Canada and various facilities in the United Kingdom. A general description of an ISD project encompasses an entombed facility which in some cases is limited to just the below-grade portion of a facility with the above-ground portion being decommissioned, packaged and disposed in a conventional manner. The entombed portions of the facility are of robust construction, generally of reinforced concrete exterior, that provides a migration barrier between internal contamination and the environment with significant internal void spaces backfilled or grouted. The scope of entombment may include ancillary equipment and structures and may contain radioactive and hazardous materials and contamination within the facility and waste imported from outside the facility. ISD is most often a permanent decommissioning end-state although in some instances (e.g. the U.K.) there may be a need to revisit the ISD facility after 300 years to effect traditional decommissioning. The defined completion (the end-state) of the decommissioned facility is therefore project-specific and in conformance with environmental approval processes and planned future activities. In all cases however, the final condition is passive, meaning that there are no requirements for ongoing operational systems or equipment within the decommissioned facility. ISD projects in the USA are presumed to be under indefinite institutional control of the U.S. Government. The regulatory framework in the U.S. is currently in place to provide assurance that the risk posed by an ISD facility is within regulatory acceptance criteria. (The framework in the U.K. and Canada is evolving towards that in the USA). Special emphasis is placed on the fact that an entombed facility is not considered a waste disposal facility; rather it is a decommissioning end-state option. ISD does not eliminate the need for effective and careful management of contaminated materials and structures, nor does it serve to abandon contaminated buildings in place. Further, ISD is safe and cost-effective for a relatively small number of facilities across the DOE Complex. This paper explores the current programmatic considerations and technical aspects of the ISD process and discuss the measures taken to ensure that an 'entombed' facility is safe for both site workers and the general population. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗