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DOE OSTI2016

Using Falcon to develop an RIA PCMI failure criteria

This paper documents the development of failure criteria for both pressurized water reactor (PWR) and boiling water reactor (BWR) fuel rod failure due to pellet cladding mechanical interaction (PCMI) during a reactivity insertion accident (RIA). This study was initiated to use cladding mechanical properties using data from separate effects tests to provide a basis to translate nuclear safety research reactor (NSRR) test data to commercial light water reactor conditions. The approach employed for the revised licensing criteria combined elements of experimental results and analytical evaluations to develop a fundamental understanding of fuel behavior during RIA events. The transient behavior of intermediate and high burnup fuel rods using well characterized RIA simulation tests was established. Cladding mechanical properties using data from separate effects tests was also developed. The Zircaloy cladding mechanical properties database of separate effects tests was used to develop the cladding integrity models. The cladding integrity models were then used in combination with the application of the Falcon fuel behavior code to translate NSRR test data to commercial light water reactor conditions. Because of the complex manner in which burnup influences fuel rod failure, it was not possible to develop a fuel rod failure threshold directly from the experimental data. An alternative approach was required that combined analytical modeling and experimental data to develop the PCMI-related contribution to the failure threshold. The analytical approach computed the radial average peak fuel enthalpy required to cause cladding failure by PCMI as a function of hydrogen content. This was accomplished by applying a critical strain energy density (CSED)-based cladding integrity model using data from total and uniform elongation tests of irradiated low-tin Zircaloy-4, ZIRLO, and Zircaloy-2 cladding materials. The CSED was developed as a function of cladding hydrogen content because hydrogen induced embrittlement is a primary effect of burnup. It was found that the failure threshold decreased with hydrogen content due to the increase in PCMI loading as a result of gap closure effects and by the decrease in cladding ductility with hydrogen accumulation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Towards a Neutronic and Thermal-Hydraulic Safety Analysis of PWR Fuel Bow with SIMULATE5 and COBRA-TF

In the 1990's, collective fuel assembly bow was observed in the Ringhals PWRs in Sweden. Back then fuel vendors reinforced the fuel assemblies' dashpot, decreased hold-down spring forces, increased guide tube wall thickness, and introduced guide tube materials with higher creep resistance in order to improve dimensional stability. Nevertheless, twenty years later in 2010, the same kind of collective bow with S and C shape patterns started to appear in most of Germany's PWRs. At this time the SIMULATE5 (S5) core simulator with its delta-gap bow model (DGA) which was originally developed for safety analysis at Ringhals, could be used as an off-the-shelf solution to assess the impact of bow on Fq and FH. However, these two quantities are of secondary importance in German PWRs in normal operation because the most restricting licensing criteria set limits on DNBRmin values. For this purpose, a method was developed to extract S5 pin powers from DGA calculations and automatically generate COBRA-TF input files (CTF) to analyze DNBR values in the cores. In the following two sections, it is shown how the S5 model for DGA calculations has been verified with independent MCNP6 calculations in order to check its predictions on pin power changes for delta gaps up to 1 cm in size and how the existing licensed DNBR calculation methodology has been updated to bowed fuel assembly situations with the help of CTF. Finally, the applicability of the Doroschuk CHF correlation, which is the licensing standard in most German PWRs, is discussed. For the fuel bow patterns considered, it is assumed that only the gap between fuel assemblies change and that the pitch between fuel rods inside the assemblies remains largely unchanged. This approximation can be justified by the visual inspections conducted at the plants during the outages. Every calculation of the pin-power impact of delta-gaps in the core under operating conditions is only as good as the knowledge of the bow pattern. Measurements of fuel assembly bow at cold conditions have been made for some German PWRs for all the fuel assemblies in the core during plant outages. Together with methods developed by the fuel assembly vendors, a best-estimate bow pattern during operation could be reconstructed. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Hazards and Probabilistic Risk Assessments of Advanced Nuclear Reactors Coupled with Industrial Facilities

This report provides a roadmap and tool kit for site specific risk assessments across a broad range of industrial customers co-located with advanced nuclear power plants (ANPP) that are not currently built and operating in the U.S. This report builds upon the body of work sponsored by the Department of Energy (DOE) Integrated Energy Systems Pathway that has produced industrial requirements studies and techno-economic assessments on the topics of feasibility of ANPP supported industrial processes. This report also leverages the DOE Light Water Reactor Sustainability (LWRS) program that has presented hazards assessment and generic probabilistic risk assessments (PRAs) for the addition of a heat extraction system (HES) to light-water reactors (LWRs) co-located with hydrogen production facilities. Many of the hazard assessments and risk assessments performed for the LWRS report are agnostic to whether the nuclear reactor is an ANPP or were adapted to the ANPP focus. The report performs hazards assessments to include industrial facilities: an oil refinery, a methanol plant, a synthetic fuel (synfuel) plant, the production of synthetic gas (syngas) as part of the methanol and synfuel plants, wood pulp and paper mills, and hydrogen production. Hydrogen production facilities are assessed in depth through prior reports in the LWRS program and the results are leveraged in this report. All these facilities are specified through industrial process and requirements research performed by national laboratories, universities, and interaction with industry. Many of the processes used in this report are pre-conceptual designs to use for decarbonization of the current technology facilities. A process of failure modes and effects analysis (what can go wrong) and accidentology (what has historically gone wrong) was used to determine the hazards presented to the nuclear power plant by the addition of the HES and the industrial customer. Chemical properties of feedstocks and products are summarized as part of the hazards assessment. Example analysis procedures are provided for each of the hazard types identified. These deterministic analyses can be used to assess adherence to licensing criteria. They can also be used to meet other safety goals like protection of the public, workers, or industrial facility equipment. A modular high temperature gas-cooled reactor (MHTGR) PRA only existing on paper was modeled and verified in modern PRA software. This will provide a tool for representative ANPP probabilistic analyses for future research.

10 SYNTHETIC FUELS↗

DOE OSTI2024

Hazards and Probabilistic Risk Assessments of a Light-Water Reactor Coupled with Industrial Facilities

This report provides a roadmap and toolkit for site-specific risk assessments across a broad range of industrial customers co-located with nuclear power plants (NPPs). This report builds upon the body of work sponsored by the Department of Energy (DOE) Light-Water Reactor Sustainability (LWRS) Flexible Plant Operation and Generation Pathway that presented hazards assessment and generic probabilistic risk assessments (PRAs) for the addition of a heat extraction system (HES) to light-water reactors co-located with hydrogen production facilities. The report expands the hazards assessments to include other industrial facilities: an oil refinery, a methanol plant, a synthetic fuel (synfuel) plant, the production of synthetic gas (syngas) as part of the methanol and synfuel plants, and wood pulp and paper mills. All these facilities are specified through industrial process and requirements research performed by national laboratories, universities, and interaction with industry. Many of the processes used in this report are pre-conceptual designs to use for decarbonization of the current technology facilities. A process of failure modes and effects analysis (what can go wrong) and accidentology (what has historically gone wrong) was used to determine the hazards presented to the NPP by the addition of the HES and the industrial customer. Chemical properties of feedstocks and products are summarized as part of the hazards assessment. Example analysis procedures are provided for each of the hazard types identified. These deterministic analyses can be used to assess adherence to licensing criteria. They can also be used to meet other safety goals like protection of the public, workers, or industrial facility equipment. The probabilistic analysis consisted of three sizes of HESs modeled in a PRA to assess the impact on the initiating events (IE) and results of the PRA. The PRA results conclude that the resulting increases in IE frequencies are below the limits required for small changes to existing NPPs under 10 CFR 50.59.

08 HYDROGEN↗

DOE OSTI2024

Cliff Retreat Rates Associated with a Low-Level Radioactive Waste Disposal Facility in Los Alamos, New Mexico, USA

We present an analysis and interpretation of potential cliff stability at a low-level waste disposal facility at Los Alamos National Laboratory, New Mexico, using cliff morphologic and fracture characteristics coupled with carbon-14 surface exposure dating. Our study is important as it directly bears on the licensing criteria for low-level radioactive waste sites. We find that future characteristic cliff failures will likely not breach disposal pits and shafts over the 1000-year minimum regulatory period. Further, we find, using a multivariate regression model, that slope angle and cliff face aspect are sub-equal in importance to predict regions of high risk of failure when combined with surface exposure ages and assuming that old exposure ages are most indicative of stability (instability) and therefore can aid decision making in final design implementation.

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DOE OSTI2024

Regulatory Treatment of Low Frequency External Events under a Risk-Informed Performance-Based Licensing Pathway: Safety Criteria Evaluation

There has been recent progress in the development of risk-informed performance-based (RIPB) design and licensing approaches, which leverage the insights gained from detailed risk assessments and utilize performance-based metrics to permit flexibility and innovation. The RIPB approaches have included the probabilistic treatment of external hazards, including low frequency events that are beyond the design basis. A preceding study conducted in FY21 noted that there are certain challenges to the probabilistic treatment of low frequency external events, which are primarily the result of uncertainty in the hazard curve and the associated plant response to rare, severe events. The current work expands on the previous study by examining the regulatory safety criteria contained within draft 10 CFR Part 53 and associated draft RIPB seismic design guidance. The evaluation identified potential challenges in demonstrating satisfaction of the safety criteria for low frequency external events, with specific difficulties associated with demonstrating compliance with cumulative safety criteria metrics, such as the quantitative health objectives (QHOs). Non-LWRs are expected to utilize the direct calculation of offsite consequence, rather than use surrogates, for comparison to the QHOs, which can be particularly challenging as the previously identified uncertainties are compounded by uncertainties in the response of the neighboring population. The current work details this evaluation and key findings. The central recommendation from this effort is that it is necessary to develop an approach for demonstrating compliance with the safety criteria in draft Part 53 that addresses the key challenges while providing flexibility to applicants.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Proposed guidance for preparing and reviewing a molten salt non-power production or utilization facility application

Development of non-power molten salt reactors (MSRs) are under consideration to further establish an MSR experience base, support the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.43(e), and provide any additional analyses needed for development of a full-scale MSR. Guidance provided in this report is based on MSRs operating with liquid fuel (i.e., fuel dissolved within a molten salt). These reactors, unless owned by the DOE or DOD, will require licensing by the US Nuclear Regulatory Commission (NRC) staff. Standard review plan (SRP) guidance for large light water reactors (LWRs) is available in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants; Light Water Reactor (LWR) Edition. However, NRC staff observed that NUREG-0800 is very cumbersome to apply to non-power reactors “because of the great differences in complexity and hazards between non-power reactors and nuclear power plants.” Therefore, a program to develop performance-based guidance applicable to non-power reactors was initiated. In 1996, NUREG-1537, Parts 1 and 2, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, was published. Part 1, the format and content guide, suggests a uniform format for presenting information in non-power reactor applications that is acceptable to the NRC staff, but conformance with the format and content is not required. Part 2, the SRP, ensures the quality and uniformity of the staff review of an application. Unfortunately, the application guidelines and SRP do not provide adequate guidance for all advanced non-LWR technologies and applications. This discrepancy eventually led to the 2012 development of interim staff guidance (ISG) for NUREG-1537, which includes criteria for describing and reviewing aqueous homogeneous reactors (AHRs). Specifically, NUREG-1537 ISG, 2012 expanded the original document to address three areas: 1. updated criteria for heterogeneous non-power reactors, 2. criteria for licensing AHRs, and 3. criteria for licensing a Part 50-licensed isotope production facility. In 2015, the US Department of Energy (DOE) opted to build on the AHR NUREG-1537 ISG experience by performing a gap analysis of the guidance that would be used to license a non-power MSR. MSRs represent one of the advanced non-LWR technologies selected by DOE for development through a multiyear cost share award with Southern Company Services. Under this Advanced Reactor Concepts 2015 (DOE Advanced Reactor Concepts [ARC] 15) award program, the DOE tasked Oak Ridge National Laboratory (ORNL) to evaluate the guidance changes that the NRC may need to consider when licensing an MSR non-power reactor. ORNL staff, with support from Boston Government Services, LLC, focused on five system-related chapters in NUREG-1537 that were considered most relevant to inform the effort that would be required for a non-power MSR applicant. ORNL documented this review in a technical report, ORNL/TM-2018/834, Proposed Guidance for Preparing and Reviewing Molten Salt Non-Power Reactor License Applications (NUREG-1537). The report was subsequently shared with industry and the NRC. The 2018 review was limited in scope, focusing on key system chapters based on the expected significance of each chapter relative to expected differences in addressing advanced non-LWR technologies, specifically non-power MSRs, compared with heterogeneously fueled non-power reactors. In the ORNL report, proposed generic adaptations were suggested for the following NUREG-1537 chapters: Chapter 4, “Reactor Description”; Chapter 5, “Reactor Coolant Systems”; Chapter 6, “Engineered Safety Features”; Chapter 9, “Auxiliary Systems”; Chapter 11, “Radiation Protection Program and Waste Management” The inclusion of Chapter 11 in the previous review effort was intended to provide guidance for categorizing the waste-handling process for an MSR operating with homogenous fuel. The introductions from Parts 1 and 2 of the 2012 NUREG-1537 ISG provide guidance for the application and review of production facilities. After a period of operation, non-power MSRs with homogenous fuel will include gaseous and soluble fission products. The gaseous fission products will be collected and held for decay in an off-gas system. There might also be an initiative to polish or filter the soluble fission products in the fuel salt by some mechanical or chemical means. The treatment and handling of fission products in the non-power MSR fuel salt and the description of this process in the safety analysis report (SAR) must be very precise to avoid the waste treatment facility being construed as a co-located special nuclear material (SNM) fuel cycle facility (see Section 2.3 of this report). Subsequent to the release of ORNL/TM-2018/834, NRC staff expressed a desire to continue the regulatory gap analysis that was begun in that report. This would provide additional clarity and information addressed in certain sections of the original report, while also providing new guidance on certain topics not addressed in the original report. This revision would benefit the NRC staff reviewing applications involving non-power MSR designs and would help developers understand how the NRC staff might approach the review of such applications. The focus of this report is to provide infrastructure support to the NRC staff for the regulatory review of non-power MSRs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Drones for Decommissioning

The U.S. Nuclear Regulatory Commission has responsibility for regulating the safe decommissioning of facilities and sites to meet the License Termination Rule in 10 Code of Federal Regulations (CFR) Part 20, Standards for Protection Against Radiation, Subpart E “Radiological Criteria for License Termination.” Decommissioning is performed in accordance with 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, as part of license termination (§50.82) and release of the facility or site for unrestricted use (§50.83). The guidance currently demonstrates the minimum requirements and necessary conditions for conducting radiological surveys by a person carrying a radiation detector(s). The Pacific Northwest National Laboratory (PNNL) evaluated the use of an unoccupied aerial vehicle (UAV) to conduct radiological surveys that could be used in decommissioning to potentially reduce time, cost, and worker safety compared to current survey methods. The objective of this project was to evaluate the performance and limitations of a UAV to support a decommissioning radiological survey and compare it to a radiological survey conducted by a human. The primary research questions of interest evaluated were: 1. Did observed UAV paths differ from human paths and, if so, how much? 2. Did survey path deviation affect survey results and, if so, how? 3. Were radiological measurements from human and UAV surveys significantly different? To answer these research questions, an experimental field was set up at PNNL’s 3440 test track, and it included radiological sources commonly surveyed during decommissioning: cobalt-60 (Co-60), cesium-137 (Cs-137), and americium-241 (Am-241). Nine check sources (three each of Am-241, Cs-137, and Co-60) with activities ranging from 3.54 µCi to 39.34 µCi were set over a path that also included an area for measuring background radiation. An Aurelia X6 UAV coupled with a GPS and lidar unit was used to conduct the radiological surveys. UAV and human surveys were conducted using two different NaI(Tl) scintillation radiation detectors (2 in. × 2 in. Ludlum, Inc. and 2 in. × 0.04 in. Alpha Spectra, Inc.) at a travel velocity of approximately 0.2 m/s at a low (15–40 cm median altitude) or high (87–105 cm median altitude) survey altitude. Since the survey velocity and altitude parameters were atypical for normal UAV operations, testing was done prior to conducting the radiological surveys to establish airworthiness, evaluate the navigation system, and establish flight control. Human and UAV surveys were paired according to the detector type and altitude regime to compare the survey data. The results of this proof-of-concept research determined that the UAV and human surveys followed similar survey paths and detected the radiological sources with no significant statistical difference (in 33 out of 36 surveys). However, further research is needed prior to deploying UAVs for decommissioning surveys.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2024

Silicon Carbide Composite Cladding Qualification

This report examines the process to qualify a new fuel-cladding type focusing on the data requirements and types of irradiation testing necessary to meet the regulatory burden and related guidance applicable to fuel qualification of accident-tolerant fuel. The report considers the use of accelerated fuel qualification techniques and lead-test specimen programs that may shorten the timeline for qualifying fuel for use in a nuclear reactor at the desired burnup. The assessment framework particularly emphasizes the identification of key fuel manufacturing parameters, the specification of a fuel performance envelope to inform testing requirements, the use of evaluation models in the fuel qualification process, and the assessment of the experimental data used to develop and validate the empirical models and empirical safety criteria.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Standard review plan for evaluating nuclear power reactor license termination plans

This standard review plan (SRP) guides the staff of the U.S. Nuclear Regulatory Commission (NRC) in conducting safety reviews of license termination plans (LTPs). The principal purpose of this SRP is to ensure the quality and uniformity of NRC staff reviews and to present a well-defined base from which to evaluate the requirements for terminating the license of a nuclear power plant. It is also the purpose of this SRP to make information about regulatory matters widely available, so that interested members of the public and the nuclear industry can gain a better understanding of the NRC staff's review process. Specific guidance for licensees is included in Regulatory Guide 1.179, "Standard Format and Content of License Termination Plans for Nuclear Power Reactors." Licensees may also use this SRP to assist in developing an LTP. The SRP presents the acceptance criteria for all areas of review for license termination, and identifies the matters to be reviewed, the basis for the review, and the conclusions that are sought. For example, NUREG-1757, "Consolidated Decommissioning Guidance" offers additional guidance on how to demonstrate compliance with the unrestricted release, restricted release, and alternative criteria for license termination, in accordance with Title 10 of the Code of Federal Regulations (10 CFR), Part 20, "Standards for Protection against Radiation," Subpart E, "Radiological Criteria for License Termination." To avoid duplication of information, this SRP references applicable sections of NUREG-1757 for additional guidance. In addition, the NRC has issued supplemental guidance that will subsequently be consolidated into NUREG-1757 to address partial site release before LTP approval, and compliance with 10 CFR Part 20, Subpart E. It should be noted that the memorandum of understanding (MOU) between the Environmental Protection Agency and NRC may contain consultation triggers for some radionuclides that may be lower than the NUREG-1757 Vol. 2 Revision 1, Appendix H screening values.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Modernization of Technical Requirements for Licensing of Advanced Non-Light Water Reactors: Safety Classification and Performance Criteria for Structures, Systems, and Components

This report, “Modernization of Technical Requirements for Licensing of Advanced Non-Light Water Reactors: Safety Classification and Performance Criteria for Structures, Systems and Components,” represents a key element in the development of a framework for the efficient licensing of advanced non-Light Water Reactors (non-LWRs). It is the result of a Licensing Modernization Project (LMP) led by Southern Company and cost-shared by the United States Department of Energy (DOE).

42 ENGINEERING↗

DOE OSTI2019

Modernization of Technical Requirements for Licensing of Advanced Non-Light Water Reactors: Safety Classification and Performance Criteria for Structures, Systems, and Components

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

NASA NTRS2009

Web-Based Programs Assess Cognitive Fitness

The National Space Biomedical Research Institute, based in Houston and funded by NASA, began funding research for Harvard University researchers to design Palm software to help astronauts monitor and assess their cognitive functioning. The MiniCog Rapid Assessment Battery (MRAB) was licensed by the Criteria Corporation in Los Angeles and adapted for Web-based employment testing. The test battery assesses nine different cognitive functions and can gauge the effect of stress-related deficits, such as fatigue, on various tasks. The MRAB can be used not only for pre-employment testing but also for repeat administrations to measure day-to-day job readiness in professions where alertness is critical.

Source record↗

DOE OSTI2016

A Comparison of the U.S. Nuclear Regulatory Commission Technical Evaluation for the Zion and La Crosse License Termination Plans with a Focus on Release Criteria

Zion Nuclear Power Station (ZNPS) and the La Crosse Boiling Water Reactor (LACBWR) sites are both undergoing decommissioning. Although these two sites are different in terms of size, design, and operational history, the approach for defining release criteria for decommissioning at each site is similar. ZNPS consists of two (Units 1 and 2) Pressurized Water Reactors (PWR), each approximately 1100 MWe. ZNPS is located near the city of Zion in northeast Illinois on the west shore of Lake Michigan. The two units were operated from 1973 to 1998. LACBWR was an AEC Demonstration Reactor (50 MWe) which operated from 1967 to 1987. The LACBWR site is in Vernon County, Wisconsin and is shared with an operational fossil plant. EnergySolutions is managing the decommissioning of both these sites under the separate entities ZionSolutions and LaCrosseSolutions. The U.S. Nuclear Regulatory Commission (NRC) transferred the Zion Facility Operating License from Exelon Generation Company, LLC (Exelon) to ZionSolutions on September 1, 2010. A similar license transfer from Dairyland Power Cooperative to LaCrosseSolutions is planned for the second quarter of 2016. On December 19, 2014, ZionSolutions submitted the ZNPS License Termination Plan (LTP). It is expected that the LTP for the La Crosse BWR will be submitted to the NRC in early 2016. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

Panel Session 57: US Nuclear Power Plant Waste Management - US LLW Disposal Issues

This panel focused on senior utility managers discussing specific disposal site issues with the disposal site operators. Representatives from each of the operating commercial disposal sites began by presenting an update of their facility, license, and waste acceptance criteria (WAC) followed by a panel discussion in response to audience questions. Panelists with presentations: Current and Emerging US NRC Regulatory Topics (Joe Weismann); WCS Update (Dan Burns); EnergySolutions Clive Facility Support for US LLW Disposal Issues (Vern C. Rogers)

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DOE OSTI2019

Panel Session 56: Nuclear Power Plant Waste Management - US LLW Disposal Issues

This panel focused on specific disposal site issues with the disposal site operators. Representatives from each of the operating commercial disposal sites presented an update of their facility, license, and waste acceptance criteria (WAC) followed by a panel discussion in response to audience questions. WM attendees that ship their radioactive waste to one or more of these commercial disposal sites attended this panel session. Presentations proposed include: 1 - US Nuclear Power Plant Waste Management - US LLW Disposal Issues (Joe Weismann, US Ecology (USA)); 2 - Waste Control Specialists Update 2019 (Dan Burns, Waste Control Specialists LLC (USA)); 3 - Barnwell LLRW Disposal Facility (Mark Lewis, EnergySolutions (USA), Benjamin Smith, Chem Nuclear Systems (USA)); 4 - EnergySolutions Clive Treatment and Disposal Facility Update (Bret Rogers, EnergySolutions (USA))

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Advanced Reactor Regulatory Development Status Report for FY 2018

This report provides a summary status of various INL-supported activities leading to the Nuclear Regulatory Commission (NRC) publication of regulatory guidance for advanced (non-light water reactor) designs. This regulatory guidance includes new design criteria designated as advanced reactor design criteria as proposed by a U.S. Department of Energy (DOE) national laboratory team to the NRC. The advanced reactor design criteria guidance was developed in response to requests from industry. The guidance is intended to provide a basis from which future advanced reactor developers and license applicants can build principle design criteria to address regulatory requirements in their specific license applications. In addition, this report provides the current status for a number of advanced non-light water reactor (non-LWR) regulatory framework development activities, including functional containment performance criteria, advanced reactor codes and standards development, and the Regulatory Technology Development Plan.

42 ENGINEERING↗

DOE OSTI2018

Radioactive Waste Disposal Strategies within the US Army Corps of Engineers FUSRAP - 18328

The U.S. Army Corps of Engineers' (USACE) highest priority is the protection of human health and the environment and public and worker safety during the cleanup, transportation, and disposal of Formerly Utilized Sites Remedial Action Program (FUSRAP) material. In addition to following all appropriate federal and state rules and regulations, the USACE has its own safety program that ensures that we are being protective. One aspect of the USACE FUSRAP is the radioactive waste disposal strategies that we must consider. The USACE will only dispose of FUSRAP material in a facility that is licenses/permitted to accept it. The estimated exposures to facility workers and members of the public from the permitted disposal of FUSRAP wastes is evaluated as part of this process. The USACE uses the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) process to study and remediate FUSRAP sites. An evaluation is performed to decide whether it is best to manage the waste on-site or off-site and the requirements for disposal. The action selected must be protective of human health and the environment and must comply with regulations. When making the decision, there are several other factors taken into consideration including cost of the disposal and whether the public and state agencies support the action. Prior to determining disposal options, wastes must first be characterized. This involves an evaluation of both site history and analytical data. Determining the proper disposal requirements and subsequent disposal option involves: Characterizing the waste as stated above. This is important because there are several regulatory agencies which may oversee management of the waste. The US Nuclear Regulatory Commission, the US Environmental Protection Agency (US EPA), the US Department of Transportation, and state hazardous waste and/or radiation protection agencies may all have varying degrees of involvement. Identifying disposal requirements. Certain wastes must be disposed at licensed radioactive waste disposal facilities. Wastes containing US EPA regulated hazardous waste must be disposed at permitted hazardous waste disposal facilities. Certain wastes must be disposed at facilities licensed/permitted for both radioactive and hazardous waste. Some lower activity wastes may be accepted for disposal at other types of landfills. Considering acceptance criteria of the disposal facility. Radionuclides and/or hazardous constituents in the waste, as well as physical and chemical form of the waste, are evaluated to ensure they meet the acceptance criteria of the receiving facility's license and/or permit. If they don't, another disposal location must be used. Considering the accessibility and location of the disposal facility. Some facilities are able to accept shipments only by truck and others can accept material via railcar as well. The distance to a disposal facility impacts the risk of transportation accidents and efficient use of project resources such as the number of trucks required to perform the transportation. These factors are also important in evaluating whether material can be cost effectively shipped to a disposal facility. Evaluating the compliance status of the receiving facility. It is not assumed that just because a facility has a license or permit that it is acceptable for the USACE' waste. Prior to selecting a facility for disposal of FUSRAP waste, a site visit, audit and consultation with the US EPA is required to determine whether there are any known relevant violations at the facility. Facilities that have violations that negatively impact the USACE or US EPA approval of the facility are not used for disposal of FUSRAP waste. Once a facility is selected. The USACE procures contracts, or utilizes existing contracts, to facilitate the transportation and disposal action. Additionally, the USACE continues to consult with US EPA routinely to confirm the compliance status of the facility. Considering state acceptance of wastes to a facility. The USACE seeks to obtain written authorization from the facility state regulator indicating the action is compliant with existing licenses, permits, laws, and regulations. Each of these factors is considered by the USACE in selecting a FUSRAP disposal location. The goal is to provide safe, cost effective disposal which is protective of human health and the environment and which complies with all regulatory requirements. These steps all feed in to the disposal strategy consideration that USACE must go through for our FUSRAP waste streams. There are a number of challenges in the process to include In order to achieve the Congressional mandate of completing cleanup work under FUSRAP in a more cost effective, timely manner, while assuring protection of public health and safety. USACE must determine all disposal options that are available for the types of waste at each FUSRAP site. As a solution, completion of a disposal strategy document can help ensure this goal is met and will foster consistency across the FUSRAP. (authors)

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