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DOE OSTI2019

Laying a Foundation for Performance Assessment (PA) Model Development - 19267

Performance Assessment (PA) models are abstractions of real-world conditions. As abstractions, modelers must make a number of decisions regarding model complexity and which features, events, and processes (FEPs) will be represented. To ensure that such decisions are transparent to stakeholders, modelers should document their approach and how this approach influences conceptual model development. This paper describes one such approach. Examples are used to describe an analysis of the potential FEPs. Through the FEPs analysis, all potentially relevant FEPs can be identified. These potentially relevant FEPs are then used to develop a detailed conceptual model known as the Central Scenario, which includes a description of the assessment context, the key features of the site and the closure system, and the expected processes of the system. From this Central Scenario conceptual model, mathematical models can be developed and implemented within modeling software and used to demonstrate compliance to PA performance objectives. FEPs which are beyond the scope of the Central Scenario are identified and used as the basis for developing alternative conceptual models (ACMs) or sensitivity analyses. Once these models and modeling cases are defined, the necessary inputs can be identified and defined. (author)

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2019

Development and Demonstration of Monitoring Strategies and Technologies -Modern2020 Project - 19317

The Modern2020 project aims at providing the means for developing and implementing an effective and efficient repository operational monitoring program, taking into account the requirements of specific national programs. The work allows advanced national radioactive waste disposal programs to design monitoring systems suitable for deployment when repositories start operating in the next decade and supports less developed programs and other stakeholders by illustrating how the national context can be taken into account in designing dedicated monitoring programs tailored to their national needs. The work has been established to understand what should be monitored within the frame of the wider safety cases and to provide methodology on how monitoring information can be used to support decision-making and to plan for responding to monitoring results. Several developments on innovative repository monitoring techniques (wireless data transmission, alternative power supply sources, new sensors, geophysical methods) has been done from the proof of feasibility stage to the technology development and demonstration phase. Innovative technical solutions facilitate the integration and flexibility of required monitoring components to ease the final implementation and adaptation of the monitoring system. Full-scale in-situ demonstrations of innovative monitoring techniques were done in way to evaluate the performance of the state-of-the-art, the innovative techniques and their comparison with conventional ones. Finally, we will introduce the experiences of the local public stakeholders that have been involved in the Modern2020 project by focusing on the participation in practice of the Belgian, French, Finnish and Swedish local stakeholders, This paper provides an overview of the European Commission Horizon2020 Modern2020 project's main goals and progress to date. (authors)

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DOE OSTI2019

International Standards as Critical Enablers to Development and Diffusion of D and D Technologies - 19372

The U.S. Department of Energy's Office of Environmental Management (DOE EM) has taken a leading role in investing in the research and development (R and D) of critical technologies to support the safe and efficient decommissioning of legacy nuclear facilities. In order to facilitate the complex-wide deployment and adoption of those technologies from the laboratory to end users, and to maximize return on investment of government-sponsored R and D projects, DOE EM proactively solicits and evaluates initiatives intended to enhance achievement of these goals. One such case study emerged and involved the successful deployment of an intumescent fixative technology in support of the SRS 235-F PuFF Facility Risk Reduction Program. A detailed analysis confirmed that a critical enabler and key contributor to this effort was a deliberate, methodical approach to leverage an international, consensus-based standards organization from the onset. There is a substantial body of literature that supports the crucial role of standards in technology programs across all phases of development - from concept, to deployment, to large-scale diffusion - in other industries. Formal standards are the result of a consensual negotiation process carried out by producers, regulators, end users and other interested stakeholders in a voluntary process within standardization organizations (WTO, 2011). Therefore, standard setting is often viewed as a self-regulatory process (Gupta and Lad, 1983), and can often yield better acceptance of technologies in certain industries, particularly if those standards are further referenced in government directives and regulations. This analysis connects several theoretical concepts: information asymmetry, risk adverse markets, and technology diffusion. It is complimented by some initial, rudimentary survey results in a burgeoning academic study conducted by the author, and reinforced by the aforementioned case study on the deployment of a fixative technology at the SRS 235-F PuFF Facility. The importance of managing a holistic technology development program that focuses on not just the technical aspects of the technology itself, but the critical enablers as well, is emphasized. This effort has direct application to potentially enhancing DOE EM technology development programs across the complex. (authors)

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DOE OSTI2019

Panel Session 83: Development of the DOE Mercury Management Strategy

This panel discussion addressed the current efforts to develop an integrated strategy for managing the mercury (Hg) challenges across the DOE complex, including challenges in remediation of Hg contamination and related ongoing or planned technology development efforts across the EM complex. Panel members discussed opportunities for collaboration in defining and implementing a mercury management strategy for the various sites, national laboratories, and contractors and for all to provide technical information for DOE to develop an integrated, cost effective Hg management strategy. Emphasis of the panel was placed on challenges at the Oak Ridge Reservation and the Savannah River Site. Opening remarks were made by Ming Zhu. Ming stated that the Academy of Sciences recently released a report on the challenges of addressing Hg in the environment; the report leveraged recent remedial efforts. Presentations proposed include: 1 - Approach to Addressing Mercury at Y-12 (Brian Henry, US DOE OREM (USA)); 2 - Mercury Management Strategy (Renee Echols, FireWater Associates, LLC (USA)); 3 - Integration of Science to Address Mercury Clean-up Challenge (Eric Pierce, Oak Ridge National Laboratory (USA)); 4 - Mercury Related Research and Development by CRESP (Kevin Brown, Vanderbilt University - CRESP (USA)); 5 - Lessons Learned from D and D of West COLEX (Kent Fortenberry, UCOR (USA)); 6 - Mercury in the SRS Liquid Waste System (Richard Edwards, Savannah River Remediation (USA)); 7 - TDD Program on Mercury at Savannah River (Bill Wilmarth, Savannah River National Laboratory (USA))

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DOE OSTI2019

Development of a Testbed for Pipeline Flushing - 19409

Pipelines which carry radioactive waste within the DOE complex should be properly flushed, particularly in the cases where stationary or moving beds of solid sediment occur, or lines are prone to hydrogen gas buildup. Present guidelines establish a minimum for the amount of water used for post-transfer flushing operations to achieve satisfactory cleaning of pipelines. However, further studies are needed to limit the amount of water used for flushing operations, thus minimizing the consequent downstream waste. These efforts are significantly helpful to DOE waste remediation sites by preserving tank storage, preventing additional waste processing, and minimizing dilution and changes in waste chemistry. An experimental pipe loop designed for study of non-Newtonian slurry flushing is currently being developed at the Florida International University (FIU). The testbed is designed to investigate parameters that effect the efficiency of flushing operations. The objective is to find flush velocity values/modes which lead to satisfactory cleaning of transport lines with a minimum amount of water usage. The developed methodologies will be streamlined according to the demands from both Hanford and Savannah River sites. A test matrix containing parameters such as pipeline length, flow rate, and simulant composition (solid concentration, rheology, particle size and density) has been developed. In the initial phase of testing, material composition and loop length will be fixed to a mixture of 17 wt% kaolin in water (to attain density, viscosity, and yield stress in the applicable range to operations at Savannah River and Hanford sites) and 165 feet, respectively. This matrix provides a parametric study on the flush velocity. Prescribed velocities in this test matrix comply with the maximum and minimum recommended flush velocity values by the flushing guidelines for Hanford waste (RPP-RPT-59600, Rev. 0, SRNL-STI-2015-00014, TFC-ENG-STD-26, REV C-1). Efforts will focus on creating repeatable sediment beds inside the pipeline in both fully-flooded and gravity-drained systems. Bed characterization (i.e., measurement of sediment height, solids concentration, and rheology) will be conducted to ensure consistency of initial conditions between tests. A combination of Pulse-Echo technique, Coriolis meter, and optical inspections will be used to accomplish initial bed characterization (ensuring consistency between tests), control/monitor density variations (initial bed creation, flush velocity control, and post-flush evaluations), and serve as reference (validation of PulseEcho and pipe surface cleanness evaluation), respectively. This paper will present the details of the development of the test loop and the test matrix. (authors)

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DOE OSTI2019

Development of a Comprehensive and Centralized Waste Management System - 19412

Information Systems Laboratories (ISL) has designed a new single, integrated, waste management and control system to administer all aspects of nuclear waste management. The overriding goal of the initiative is to provide federal, state, and local agencies with operational awareness, management, and programmatic support for all functions related to nuclear waste management. Over decades of waste management federal agencies have sponsored or directly paid for the development, operation, and control of multiple systems to track and control its waste - with varied success. Multiple contractors have generated disparate control systems/databases and literally mountains of data. Past approaches have not provided a single, integrated, real time data management system. ISL's analysis of past failed efforts has reveal four common symptoms: - Control/management systems that are not compatible; - Programmatic costs that increase with systems use and re-work; - Systems focused on specific sets of data; - Inaccurate or outdated data sets. An integrated waste management and control system that collects and displays real-time, accurate inventory, waste container certification, transportation certification, emplacement, and shipment related data improve operational awareness and overcomes these limitations. Using its vast knowledge and experience in developing numerous waste tracking and management systems, ISL has leveraged its own in-house knowledge and gathered requests from stakeholders all across the industry to design a system that meets the needs of users at all levels of the nuclear waste management process. Using the highly successful and widely used Waste Data System (WDS) designed for the Waste Isolation Pilot Plant as a core component, ISL has designed a system that greatly expands on the functionality of the WDS to exceed its current focus of management of Transuranic Waste shipments. The new system is designed to provide the ability to track a wide range of waste types from Transuranic, High Level, Low Level, Mixed, and any nuclear related material. The main focus of the system is to provide federal regulators a comprehensive and real-time oversight into the handling of all nuclear waste materials. Oversight activities include, but are not limited to, strategy development, projections, metrics, incentive development, tracking, budget review, government cost estimating, and integrated schedules. The major components of the system are summarized below: - The Centralized Waste Management System (CWMS), which will include assimilating and validating known sources of data into the CWMS. - The programmatic reporting support component, which will handle queries from regulators, track compliance with regulatory requirements, and maintain the dashboard for federal management. - Both the CWMS and the programmatic reporting support component will use a standard set of business rules that govern system operations and allow site and program specific needs and limits for all waste programs and types to be captured and maintained. This presentation will provide details of each component of the system detailing its use, interfaces with other components, and value to stakeholders. (authors)

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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)

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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)

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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)

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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)

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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)

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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)

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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)

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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)

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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↗