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DOE OSTI2019

Developments in the Regulatory Framework of Nuclear Waste Management in Germany - Licensing Procedures for storage of SNF and Public Involvement - 19612

According to article 6 of the German Atomic Energy Act (AtG), storage of Spent Nuclear Fuel requires a license. The competent licensing authority in this field is the Federal Office for Safety of Nuclear Waste Management (BfE). According to the Act for selecting a repository site for heat-generating radioactive waste (HAW) (Site Selection Act - StandAG) dating from 2017, BfE has been assigned the task of developing, conducting, and evaluating public participation procedures in the process of site search and selection for a deep geological repository for radioactive waste and SNF. Article 6 AtG applies to interim storage of SNF. If the legally defined licensing conditions by article 6 (2) AtG are met, the storage license applied for has to be granted. For on-site facilities, interim storage has been limited to 40 years. The licensing procedures for interim storage of SNF are carried out in accordance with the general administrative procedural rules. In certain cases due to the Environmental Impact Assessment Act (EIA Act), an EIA has to be carried out. In these cases, a formalized licensing procedure including an involvement of the public is conducted. The Nuclear Licensing Procedure Ordinance (AtVfV) originating from 1977 and reformulated in 1995 defines formalized rules for the public participation procedure which have to be followed strictly. These include a public hearing. The risen objections are evaluated by the licensing authority and taken account of in the licensing decision. The process of repository site selection defined by the StandAG exceeds these formalized procedures significantly: The objective of this process is to achieve a broad social consensus about the site selection procedure and about a repository site that will offer the best possible safety for one million years. Therefore, the process has to be particularly transparent, and involvement of the public - including newly developed forms of public involvement - is an essential part of it. BfE as competent authority in both the fields of repository site search and selection and licensing of interim storage of SNF has recently developed new strategies of involving the public concerning interim storage of SNF. Supplementing the standard licensing procedures, BfE presently informs and involves the public about topical issues regarding the present state and recent developments in the field of interim storage of SNF. Current central topics are for example the return of HAW and MAW from reprocessing of SNF in France and UK as well as perspective longer interim storage periods on the interface with final disposal. Regarding longer interim storage, host communities of interim storage facilities are concerned that interim storage might be extended indefinitely. Additionally, there is a strong fear of transports of radioactive waste and SNF from one place to another. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Developing a U.S. High Performance Research Reactor Transportation Road-map - 19674

The U.S. National Nuclear Security Administration Office of Reactor Conversion through the U.S. High Performance Research Reactor (USHPRR) Project is developing and qualifying a new Low Enriched Uranium (LEU) fuel to convert USHPRRs from the use of highly enriched uranium fuel to LEU fuel. The USHPRR Project is forecast to complete in 2030's. During the Project's life cycle, it will make hundreds of radiological shipments. These shipments will include material of different sizes and radiological characteristics and involve many facilities. These shipments must be integrated with the Project's detailed schedule to ensure success of the Project. The USHPRR Project has the complex challenge to develop a Packaging and Transportation (P and T) Road-map to coordinate the multitude of packaging and shipment activities. A comprehensive packaging and shipping matrix is being developed to assist in meeting this challenge and ensure the USHPRR shipments can be made on schedule. The vision for the matrix is for it to be a bridge between the schedule and the P and T Road-map, and for it to be the one stop place for USHPRR shipment information. An interactive approach to the P and T Road-map and the comprehensive shipping matrix provides an effective strategy to execute the USHPPR Project today and through its life cycle. This paper provides an overview of the development of the USHPRR P and T Road-map and the comprehensive shipping matrix and their benefits to the USHPRR Project. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Development of Concentration Ratios and Transfer Factors for Human and Ecological Probabilistic Performance Assessments - 19781

Human and ecological risk assessments utilize concentration ratios and transfer factors to calculate contaminant concentrations in plant and animal tissues in equilibrium with soil or water (concentration ratios) or based on an animal's contaminant intake from various media (transfer factors). These ratios and factors are used to evaluate human dose via food ingestion pathways and to evaluate ecological dose to an organism. In a probabilistic performance assessment, concentration ratio and transfer factor distributions are developed for every element, and these distributions are applied across the spatial and temporal domain of the assessment. As a result, it is necessary to scale the underlying data distributions by applying the central limit theorem to develop appropriate distributions of the mean for the temporal and spatial scale of the model. Herein we discuss the methodologies for identifying and ranking references, and how to develop statistical distributions of concentration ratios and transfer factors for probabilistic performance assessment models. We also discuss factors that could be pertinent to differences when selecting references and developing distributions for human versus ecological risk assessment scenarios. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

ANDRA's Underground Research Laboratory in Bure: Major Role in the Cigeo Development - 20005

The Industrial Center for Geological Disposal, also called Cigeo, is the deep geological disposal facility project developed by Andra since 1991. It is intended for the final disposal of High-Level Waste (HLW) and Intermediate Level Waste-Long Lived (ILW-LL) generated in France by the nuclear industry. Cigeo is located in the east of France (Meuse/Haute-Marne site). The disposal will be implemented in a 140 to 160 m thick clay layer at about 500 m depth. The license application file will be submitted in 2020. If the license is granted, the construction of the pilot phase of Cigeo (ramp, shafts, drifts, initial disposal vaults) could start in 2025 at the earliest. Since 2000, the development of the safety case of Cigeo for post closure has been supported by a three-stage construction, design, and scientific and technological experiment program performed in the French Underground Research Laboratory (URL) in Bure. These stages of the URL activities supported the iterative interactions between the knowledge acquired by scientific and technological R and D program, the design process and the safety assessment. The stages focused on assessing the suitability of the disposal concept, providing the basis for safety options and construction design, and preparing for licensing using large scale demonstrations. Recently, Andra launched the fourth development stage of the URL to implement a new set of technological experiments aiming at consolidating the design options of disposal cells and galleries for the pilot phase of Cigeo and at assessing monitoring technologies. Recently, Andra launched the fourth development stage of the URL to implement a new set of technological experiments aiming at consolidating the design of the pilot phase of Cigeo and at assessing design options and monitoring technologies. Removal of a segment ring, construction of an X drift crossing, improved construction techniques for HLW vaults, and construction of an ILW-LL prototype disposal vault are example activities during the fourth phase. In the future, the URL in Bure will remain a unique location to carry out research on promising technical solutions, to reduce Cigeo's construction and operation risks and strengthen the long-term safety assessment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Panel Session 69: Ensuring Strong Return on Investment for DOE EM Technology Development: Global Views

This panel focused on DOE and Non-US Technology Development Projects discussing how they approach tech development to maximize return on investment (ROI). Successful examples were discussed. After the individual panelist's presentations, the three panelists were joined by Kurt Gerdes in discussing the genesis of this panel and why it is important. Panelists with presentations: Ensuring a Strong Return on Investment for DOE EM Technology Development (Anthony Banford) DOE Technology Development Projects that Provided a Strong Return on Investment (ROI (Mark Lesinski) Australian Nuclear Science and Technology Organisation (Gerry Triani) Decommissioning Situation of Nuclear Power Plant in Japan (Teruyuki Hirai)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Developing and Implementing New Waste Chemistry Controls for Hanford's Double-Shell Tanks - 20041

Waste chemistry controls for the Hanford double-shell tanks (DSTs) were established in the 1980's in response to tank failures caused by stress corrosion cracking (SCC) at the Savannah River Site (SRS). The controls were established based on a combination of SRS chemistry limits and corrosion testing results. Conditions in the Hanford DSTs have changed over time, with the temperature in the majority of the DSTs dropping below 50 deg. C. Testing indicated that the original chemistry control specification and SCC testing performed in 2010 did not protect against pitting corrosion mechanisms - what is believed to have caused the failure of tank 241-AY-102. New waste chemistry controls are necessary to protect the DSTs from both SCC and pitting corrosion. Savannah River National Laboratory (SRNL), with input from the Tank Integrity Expert Panel (TIEP) Corrosion Subgroup (CSG), performed a statistically based investigation of nitrate and halide ion induced pitting corrosion. The investigation was intended to develop a comprehensive waste chemistry envelope that minimizes the risk of both SCC and pitting caused by halide and nitrate ions. The waste chemistry envelope needed to be robust enough to address future waste retrieval and process changes which could significantly change the waste composition of the DSTs. The experimental design allowed the statistically significant ions and their effect on pitting corrosion to be determined. SRNL developed the 'pitting factor' which is an empirical relationship between the statistically significant inhibitive species (hydroxide, and nitrite), and pitting inducing species (nitrate, chloride, and fluoride). The coefficients are weighted factors that show the degree of influence the species have on the propensity for pitting. New waste chemistry control limits were recommended by SRNL, and concurred with by the TIEP CSG, that incorporated the pitting factor along with other limits. The controls developed to reduce the risk of pitting corrosion were evaluated to determine if they also protected against SCC. Hundreds of test results were reviewed and compared against the proposed limits. With the exception of one test result, instances where cracking was observed corresponded to chemistry that was outside of the proposed limits - indicating that the chemistry controls developed for pitting also adequately addressed SCC concerns. Existing tank composition estimates were evaluated against the new chemistry controls - three current tank chemistries are suspect, and may not meet the new chemistry control requirements. These included the interstitial liquid (liquid associated with solids) of DSTs 241-AN-102, 241-AN-106, 241-AN-107, and 241-AY-101. Measures are ongoing to evaluate the tanks with the suspect chemistry. These include varying combinations of: performing additional evaluations (history of use, modeling); DST specific laboratory testing; and, pursuing core samples of the DSTs. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Tool for Developing Integrated Strategies for Decontamination and Waste Management - 20291

Management of waste from chemical, biological, and radiological incidents is complicated by the fact that decontamination decisions have a profound impact on the characteristics of resulting waste streams. Wide-area biological and radiological incidents could produce massive quantities of waste that may need to be treated on-site and sent for subsequent disposal as non-contaminated materials, or else be directly disposed of as contaminated materials. The EPA has developed the Waste Estimation Support Tool (WEST) for characterizing and quantifying biological and radiological waste that may be generated from decontamination efforts. This paper focuses on the WEST's uses for radiological incidents. WEST combines Geographic Information System (GIS)-based analysis of externally-supplied plume data, infrastructure databases derived from the Federal Emergency Management Agency's (FEMA's) Hazus tool [1], and satellite imagery surface recognition algorithms to combine the composition and square footage of the buildings in the plume with estimates of the materials between the buildings in the plume. The resulting GIS data files are then imported into a Microsoft Access database application, where they are combined with information about the nature and concentration of contaminants, and then subjected to decontamination strategies for different contaminated surfaces. The tool provides estimates of the type and quantities of potential wastes resulting from simulated decontamination and/or demolition activities and includes estimates of the remaining contamination levels including residual contamination contained within each waste stream. Estimates are presented at several levels of detail, allowing users to obtain needed data at the desired resolution. These include estimates for the total affected area, estimates by contamination zone, estimates by decontamination method(s), and estimates by building type (occupancy classification). EPA is currently developing the next version of WEST which will include several substantial enhancements. The most significant improvement for the next version of WEST will include the ability for users to develop contamination scenarios and waste estimates based on previously developed, readily available, and geographically specific infrastructure data. Instead of using WEST's default infrastructure data based on FEMA's Hazus tool, users will be able to import their own building data specific to the geographically affected area. This capability may substantially decrease uncertainties in the resulting waste estimates because the results will be based on actual building data (numbers of each type, square footage, building height, etc.). Two other significant enhancements will be the ability to generate waste estimates for vehicles and biomass. In addition to building debris and building decontamination waste, vehicles and biomass will likely constitute a significant percentage of the total waste which may result from wide area contamination events. This presentation will present the most recent version of WEST, which includes such considerations as affected biomass (e.g., trees), vehicles, and the ability to replace the default Hazus infrastructure databases with custom infrastructure databases. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Virtual and Augmented Realities to Support Development, Training, Deployment, and Operations of Remote Systems - 20516

Throughout the U.S. Department of Energy complex, there is a need to deploy new, unique equipment and systems for cleanup activities, waste processing, infrastructure repair, and decommissioning and demolition in radioactive and hazardous environments. The new equipment often consists of custom systems created by integrating and modifying commercial technologies. In addition, the systems are deployed in unique environments due to limited access, waste properties, and the presence of previously installed equipment or infrastructure that has been left in place. The deployment of custom systems that are 'turnkey' ready for first-time installation and application has its risks. Use of test platforms and mockups for cold (i.e., non-radioactive) testing is one method to mitigate that risk: however, full-scale test mockups for operations such as waste retrieval and tank inspection can be expensive and must be modified to incorporate the uniqueness of each application. An alternative to full-scale mockups is to use three-dimensional (3D) virtual environments and virtual reality (VR) technology to develop interactive virtual test beds. These can be used throughout the down-selection, development, system integration, operational planning, operator training, and mission execution phases of a task. Virtual test beds are readily reconfigured, can be used for multiple tasks, support multiple users in parallel, and can accommodate enhancements and changes not originally planned or conceived in the initial design. Virtual test beds also provide a platform for developing and testing augmented reality (AR) tools for improving operator efficiency. A VR experience has been generated for waste retrieval from a Hanford waste tank to present the potential applications of a virtual test bed. Features of the VR experience are presented, potential applications of VR and AR for remote applications are discussed, and development areas for improving future application are considered. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Mechanistic source terms for the XE-100 reactor XSTERM Code Development Approach - 18587

X-Energy is developing the Xe-100 pebble bed modular High Temperature Gas Reactor (HTGR) as an advanced nuclear power source. To achieve its safety objectives, the Xe-100 reactor relies on materials inherent characteristics and passive safety features. Key features include the inherent high temperature characteristics of TRISO-coated fuel particles, graphite moderator, and helium coolant, along with passive heat removal capability of a low-power-density core. This paper will describe the approach adopted by X-Energy to develop and validate mechanistic source terms for the Xe-100. The Mechanistic Source Term (MST) approach for the Xe-100 reactor will build upon that of the prior DOE Next Generation Nuclear Program (NGNP) (Refs. 1, 2, 3 and 4). MST development is closely related to Fuel Qualification (FQ), Licensing Basis Event (LBE) selection and Probabilistic Risk Assessment (PRA). This paper provides a brief overview of the MST road map process. Software plays a crucial role in establishing and quantifying MSTs required for licensing the Xe-100. An insight into X-Energy s comprehensive, integrated software development program for calculating time-dependent radionuclide releases from the Xe-100 over a wide range of events is provided. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

AREVA Steam Cycle High Temperature Gas-Cooled Reactor Development and Deployment Plan - 18684

The development and deployment timeline for the AREVA SC-HTGR is among the shortest of all Generation IV reactor technologies. Once demonstrated, the SC-HTGR enters an expanding marketplace beyond electricity generation. This enables industrial growth that is today solely captive to fossil fuels, primarily natural gas supply. The SC-HTGR plant will be a hedge that insulates a wide range of process industries from fossil fuel price and supply volatility in coming decades when the SC-HTGR plant design is ready for deployment. Similar to existing light water reactors, the SC-HTGR is largely immune to fuel price swings for its expected 60 year operating life cycle. Approximately 70% of the SCHTGR plant energy cost is driven by the initial capital investment with fuel being less than 20%. This is entirely the opposite of natural gas fueled power plants where about 70% of the energy cost is directly tied to the cost and availability of fuel and the potential volatility this brings with it over a plant life cycle that could span several decades. This paper details the development and deployment strategy envisioned by AREVA for the SC-HTGR plant. The development phase includes completion of the full scale reactor module design and licensing, build and demonstrate the first commercial scale module, get design certification, and launch the commercial fleet deployment. The development plan takes full advantage of a) a large body of successful pre-application licensing interactions with the NRC under the DOE NGNP project, b) ongoing experiments and test results accumulated through the TRISO particle fuel qualification and nuclear grade graphite characterization R and D activities conducted in the U.S. and internationally, and c) operation of several small scale technology demonstration facilities and large scale commercial design projects. The abundance of the historic HTGR reactor technology demonstration projects provides a high degree of confidence that sufficient technical information is available to begin FOAK plant engineering and licensing followed by commercial demonstration of the first reactor module within a next decade. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Development of a utility computer-based training program on setpoint methodology for nuclear power plants - 129

The purpose of this paper is to provide an overview of the development of the Computer- Based Training (CBT) Program for utility engineers on Setpoint Methodology for nuclear power plants. This computer based training module was prepared by the Advanced Engineering Training initiative. The Advanced Engineering Training initiative is a nuclear industry effort to develop technical computer-based training modules for targeted areas of engineering expertise. The target audience is engineers at commercial nuclear power facilities in the U. S. These modules are being developed using the systematic approach to training and are being designed to maintain industry engineering expertise and to facilitate engineering knowledge transfer. The objective of this training effort is to support the development of the next generation of site technical experts for specific topic areas. The training is one part of three key components that would facilitate knowledge transfer to the future site Subject Matter Expert (SME). The following are the major focus areas covered in the Setpoint Methodology CBT: Chapter 1 provides an overview of the role instrument setpoints perform in the defense-in-depth safety analysis for the plant as well introducing key definitions related to setpoints. Chapter 2 provides the history and current regulatory requirements related to setpoint control including the laws and guidance provided by the U.S. NRC. Chapter 3 provides describes the setpoint methodology including terms, formulae, methods of combination, graded approach and methods for drift analysis. This is a major part of the training module in covering the incremental uncertainty terms that need to be considered and how they are combined. Chapter 4 provides and overview of the role of setpoints in Technical Specifications and associated documents. Chapter 5 describes how the regulatory process, related to setpoint control, has been implemented in NRC inspections and how the major elements of the 10 CFR 50.59 reviews apply to the performance of a new or revised setpoint calculation. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Supporting resilience in nuclear power plant operations - conceptual frames for human factors development and validation - 166

This is a summary paper of conceptual developments on nuclear power plant operation by the human factors team at VTT Technical Research Centre of Finland. While drawing together years of studies, the paper also discusses recent progresses in training development. Firstly, there is a long background in work analysis and system validation. The analysis model applied provides a model for assessing and identifying 'core-task functions,' that is, basic functions that should be fulfilled within work activity. Secondly, we have evaluated the existing training practices at a Finnish plant and identified challenges that could be generally prevalent in the nuclear domain - they seem to reflect the hierarchical and safety-critical nature of NPP operation. In terms of potential solutions, we are currently testing and developing some new self-evaluation and self-confrontation methods. First and foremost, however, we discuss the common theory basis for our research approaches. On the broadest level, our development and research work reflects the view of safety as expressed by resilience engineering literature: safety is not seen merely as 'negative' lack of mistakes, but also as 'positive' capability to solve and anticipate problems. For example, it is assumed that operators' capability for situational interpretation should be supported at all times: even if the emergency procedures dictate operator activity specifically, the operators should understand the influence of the operating procedures for the plant process and system state. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Development of utility generic functional requirements for electronic work packages and computer-based procedures - 273

The Nuclear Electronic Work Packages - Enterprise Requirements (NEWPER) initiative is a step toward a vision of implementing an eWP framework that includes many types of eWPs. This will enable immediate paper-related cost savings in work management and provide a path to future labor efficiency gains through enhanced integration and process improvement in support of the Nuclear Promise (Nuclear Energy Institute 2016). The NEWPER initiative was organized by the Nuclear Information Technology Strategic Leadership (NITSL) group, which is an organization that brings together leaders from the nuclear utility industry and regulatory agencies to address issues involved with information technology used in nuclear-power utilities. NITSL strives to maintain awareness of industry information technology-related initiatives and events and communicates those events to its membership. NITSL and Light Water Reactor Sustainability Program researchers have been coordinating activities, including joint organization of NEWPER-related meetings and report development. The main goal of the NEWPER initiative was to develop a set of utility generic functional requirements for eWP systems. This set of requirements will support each utility in their process of identifying plant-specific functional and non-functional requirements. The NEWPER initiative has 140 members where the largest group of members consists of 19 commercial U.S. nuclear utilities and eleven of the most prominent vendors of eWP solutions. Through the NEWPER initiative two sets of functional requirements were developed; functional requirements for electronic work packages and functional requirements for computer-based procedures. This paper will describe the development process as well as a summary of the requirements. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Development of a Multiphase Adjoint Capability in OpenFOAM

Thermal hydraulic predictions of flow regimes within nuclear reactor cores require significant computational resources and accuracy to ensure that the core design does not violate the thermal limits of materials. Multiphase computational fluid dynamics (CFD) has become a valuable computational capability for predicting flow characteristics as they relate to applications such as nuclear reactor safety and mixing phenomena around spacer grids within fuel assemblies. Multiphase CFD requires closure relationships for both the effects of turbulence as well as the transfer of mass, momentum, and energy between phases. Understanding the sensitivity of the parameters used in closure relationships is important for improving accuracy, but performing rigorous sensitivity analysis is computationally expensive. Adjoint mathematics have been utilized for reactor physics and fluid dynamics applications with respect to verification, sensitivity analysis, adaptive mesh refinement, and optimization. The usage of adjoint methods to predict the sensitivities of a variety of closure relationship parameters within a given problem has been shown to return sensitivity responses with significantly fewer computations than normal 'large-scale' methods. As the complexity of multiphase CFD methods increases, the demand for accurate verification tools and improvements in computational efficiency also increases, making the development of adjoint capabilities an attractive option. Closure relationships that describe interphase momentum physics in multiphase CFD modelling rely heavily on prescribed constants determined from experimental data. Due to the computational efficiency of an adjoint approach, the time to derive these closure relationships could be dramatically reduced, and data assimilation techniques could be used to return probability density functions of closure models derived using experimental data. This work develops a multiphase adjoint capability for the Eulerian-Eulerian two field, two phase equations. Specifically, the adjoint capability has been developed for the sub-cooled nucleate boiling solver boilEulerFoam developed by Dr. Nam Dinh of NC State University within the open-source CFD library OpenFOAM.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Development of Human Factors Awareness for Criticality Safety Specialists at Westinghouse Nuclear Fuels - 14223

Although human factors assessment typically involves specialist expertise, it is considered that enhanced local knowledge of this area for criticality safety specialists supports the incorporation of human factors consideration into tasks such as safety assessment, design and hazard reviews, development of operational and engineered criticality controls, training, incident investigation and emergency response. At the Westinghouse Springfields Fuel Manufacturing Site in the United Kingdom, specialist human factor support has been engaged to identify key competencies and training plans for criticality safety specialists. In addition, discussions have been held with the UK's Office for Nuclear Regulation (ONR) to outline the proposal and to discuss the approach. Following completion of an internationally recognised human factors training course the learning is now being applied by criticality safety specialists. Recent examples of learning applied include: 1) Review of past events and incidents to support the consideration of operator/equipment interfaces, leading to improved designs and processes. 2) Identification of critical tasks and the level of human factor assessment that is likely to be required. To support this, guidance has been produced by the criticality section for staff and contractors performing criticality safety assessments. 3) Assessment of the 'aims and operability' of engineered and procedural controls derived as part of the criticality safety assessment process. This involves participation of plant management and operators. 4) Consideration of human factors in emergency response activities such as infrequent tasks and tasks being undertaken under time pressures etc. This has been examined in recent emergency exercises and in the development of a recent criticality exercise scenario. 5) Criticality training has been reviewed with Plant Management, focussing on the methods utilised to deliver training (e.g. classroom based, computer training, on plant walk-downs); the aims and scope of training and how understanding is validated. 6) The link between criticality safety assessment and how the operations are / will be conducted has been reviewed as part of formal and informal criticality meetings and inspections. A key aspect of this is the criticality specialist's interaction with both plant management and operational staff. The development of human factors awareness into the criticality safety specialist's competency framework continues to be developed at Westinghouse but real benefits to criticality safety are already being demonstrated by this approach. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Development of criticality risk evaluation method for fuel debris in Fukushima Dai-ichi NPS - 14373

After the accident of Fukushima Dai-ichi Nuclear Power Station, there has been still damaged nuclear fuel in the reactor buildings, which is called 'fuel debris' and contains an amount of fissile materials show that it is currently subcritical, which must be kept until finishing retrieval work. In the latest plan, retrieving fuel debris is to start in 2025. The International Research Institute for Nuclear Decommissioning, IRID, has been developing the technology for decontamination in reactor building. A developed method will be reviewed and granted by the Nuclear Regulatory Authority in Japan, NRA, before the start of the retrieving. Criticality risk information will be needed for decision making on following two subcriticality issues, -Criticality control of fuel debris to keep is subcritical until the end of retrieval work. -Safe operation for retrieving fuel debris to keep it subcritical. Because fuel debris has been out of criticality control and the key condition for its current subcriticality is unknown, it is important to obtain criticality risk information as soon as possible. Applying conventional Probabilistic Safety Assessment, PSA, to fuel debris is difficult, because it is based on the probability of becoming out of criticality control due to troubles such as equipment failure, operation mistake and so on. A new method, which has been developed by Japan Atomic Energy Agency to technically support the NRA, is designed to estimate criticality risk based on a probability distribution of k{sub eff} due to uncertainty of composition and geometry of fuel debris, which method will provide risk information for the two subcriticality issues as well as for the tolerance level of criticality risk if it is unavoidable. Developed method contains four steps: two steps of analysis of criticality risk and two steps of quantification of the risk. A procedure of risk quantification has been defined, in which systematic analysis of criticality risk can be done by making 'status matrix'. A trial analysis by using the procedure shows that some events which can initiate criticality are able to be supposed to happen during maintenance period until the beginning of retrieving fuel debris, and that analysis also indicates countermeasures should be considered as soon as possible in order to reduce criticality risk for public and workers. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Development of a process for remediation of enriched uranium coolant tanks at AWE Aldermaston - 14520

Machining processes involving Enriched Uranium (EU) at the UK Atomic Weapons Establishment have historically required the use of a liquid coolant to ensure product quality and to provide protection against hazards associated with ignition of the EU swarf in the non-inert gloved air box atmosphere. This coolant is typically provided by closed circuit re-circulating systems utilising large volume tanks to contain the bulk coolant. Small losses of EU to the coolant tank are experienced during the machining process leading to accumulation over time. This is expected and appropriate measures are put in place to control such accumulations to ensure criticality safety during the operational phase. The original design intent made no specific provision for recovery of the EU as it was not expected to be required based on the anticipated losses and design lifetime of the tanks. However, in support of ongoing operations it has become necessary to develop a process to remediate out of service coolant tanks to recover the EU and to replace the coolant and some of the end-of-life tank components. Detailed optioneering has been undertaken to develop a process that addresses all the relevant hazards, including nuclear criticality safety. Non-destructive assay techniques have been developed to help determine the inventory of EU present and consequently the level of control required during the remediation process. This has led to the selection of favourable geometry collection vessels that not only afford significant defence-in-depth against criticality hazards during the remediation work but also satisfy interim storage requirements. These vessels may also be of interest to other operators. This paper presents an overview of the history of operation of the coolant tanks and proposed remediation process and provides an overview of the criticality safety case that has been developed to support it. It is anticipated that the remediation work will be undertaken before 2019 allowing the findings and lessons learnt (including validation of non-destructive assay results on expected hold up) to be reported at the next ICNC. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Development of Shielding Benchmarks Using the Godiva IV Assembly

The Nuclear Criticality Safety Program (NCSP) is developing a shielding benchmark using the Godiva IV assembly as a source. Its present status is reviewed herein. Even eight decades into the nuclear era, substantial work remains to develop a database of shielding benchmarks to support future nuclear development. A nuclear simulation is only as good as its supporting data, inputs, and validation basis. Uncertainty in these areas is addressed using conservatism, which adds margin and, occasionally, cost. In many shielding situations, high accuracy is not necessary because additional material is not particularly expensive. After all, 1–2 cm of lead often reduces the gamma ray dose substantially. However, in certain areas, conservatism can add unnecessary cost. These areas include mobile shielding applications such as casks, ships, microreactors, and spacecraft, where weight and, thus, margin is expensive. Although these characteristics are side benefits for NCSP shielding benchmark development, the main driver is enabling more reliable placement of criticality accident alarm systems (CAASs) in nuclear material facilities, such as those dedicated to the production of advanced reactor fuels. CAAS placement relies on more than accurate data and code validation. It also relies on sufficiently accurate materials specifications, geometry specifications, and a well-defined, alarm-producing baseline accident. All these things require tacit knowledge and understanding of the problem being evaluated. Benchmarks can help ensure this understanding.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗