Evaluation of Local Enforcement of Energy Efficiency Standards and Labeling Program in China
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29 ENERGY PLANNING, POLICY, AND ECONOMY↗
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Promising Fuel Cycle Options for R&D – Results, Insights, and Future Directions
The Fuel Cycle Options (FCO) campaign in the U.S. DOE Fuel Cycle Research & Development Program conducted a detailed evaluation and screening of nuclear fuel cycles. The process for this study was described at the 2014 ICAPP meeting. This paper reports on detailed insights and questions from the results of the study. The comprehensive study identified continuous recycle in fast reactors as the most promising option, using either U/Pu or U/TRU recycle, and potentially in combination with thermal reactors, as reported at the ICAPP 2014 meeting. This paper describes the examination of the results in detail that indicated that there was essentially no difference in benefit between U/Pu and U/TRU recycle, prompting questions about the desirability of pursuing the more complex U/TRU approach given that the estimated greater challenges for development and deployment. The results will be reported from the current effort that further explores what, if any, benefits of TRU recycle (minor actinides in addition to plutonium recycle) may be in order to inform decisions on future R&D directions. The study also identified continuous recycle using thorium-based fuel cycles as potentially promising, in either fast or thermal systems, but with lesser benefit. Detailed examination of these results indicated that the lesser benefit was confined to only a few of the evaluation metrics, identifying the conditions under which thorium-based fuel cycles would be promising to pursue. For the most promising fuel cycles, the FCO is also conducting analyses on the potential transition to such fuel cycles to identify the issues, challenges, and the timing for critical decisions that would need to be made to avoid unnecessary delay in deployment, including investigation of issues such as the effects of a temporary lack of plutonium fuel resources or supporting infrastructure. These studies are placed in the context of an overall analysis approach designed to provide comprehensive information to the decision-making process.
continuous recycle↗
Scout: An Impact Analysis Tool for Building Energy-Efficiency Technologies
Evaluating the national impacts of candidate U.S. building energy-efficiency technologies has historically been difficult for organizations with large energy efficiency portfolios. In particular, normalizing results from technology-specific impact studies is time-consuming when those studies do not use comparable assumptions about the underlying building stock. To equitably evaluate its technology research, development, and deployment portfolio, the U.S. Department of Energy's Building Technologies Office has developed Scout, a software tool that quantitatively assesses the energy and CO2 impacts of building energy-efficiency measures on the national building stock. Scout efficiency measures improve upon the unit performance and/or lifetime operational costs of an equipment stock baseline that is determined from the U.S. Energy Information Administration Annual Energy Outlook (AEO). Scout measures are characterized by a market entry and exit year, unit performance level, cost, and lifetime. To evaluate measures on a consistent basis, Scout uses EnergyPlus simulation on prototype building models to translate measure performance specifications to whole-building energy savings; these savings impacts are then extended to a national scale using floor area weighting factors. Scout represents evolution in the building stock over time using AEO projections for new construction, retrofit, and equipment replacements, and competes technologies within market segments under multiple adoption scenarios. Scout and its efficiency measures are open-source, as is the EnergyPlus whole building simulation framework that is used to evaluate measure performance. The program is currently under active development and will be formally released once an initial set of measures has been analyzed and reviewed.
32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
Insights from PPL's DERMS Implementation
PPL Electric Utilities’ Keystone Solar Future Project (KSFP) is part of the U.S. Department of Energy’s Office of Energy Efficiency and Renewables, Energy Enabling Extreme Real-time Grid Integration of Solar Energy (ENERGISE) grant. The three-year project, beginning in October, 2017, funded an innovative pilot program to develop and evaluate a technology platform and framework for managing high levels of distributed energy resources (DER), and improving future DER interconnection. The KSFP aimed to bridge gaps between existing and future technology, including Distributed Energy Resource Management Systems (DERMS), by monitoring, managing, and optimizing a system with high penetration of solar and other DER. Specifically, the project worked to make Demand Management Systems (DMS) applications, devices, and real-time communications technology DER-aware to allow for advanced applications for utility operations. Here, by installing and implementing a DERMS, project designers transformed how DERs interact with PPL’s distribution system, offering learnings and results worth sharing industry-wide.
Leader, Jared↗
Recent Developments in DOE FUSRAP - 13014
The DOE Office of Legacy Management assumed responsibility for the DOE FUSRAP in 2003. Since then, DOE has evaluated existing guidance and program needs to ensure that the program will remain effective in maintaining protectiveness at remediated FUSRAP sites. DOE has identified crucial elements that must be addressed to meet this goal. Knowledge of the sites and the program must be preserved and accessible to future custodians. Program processes must be defined and coordinated with other agencies. Long-term surveillance and maintenance (LTS and M) requirements for the sites must be based on human health risk and regulatory compliance, and those requirements must be well defined. Useful and accurate program information must be available to stakeholders. DOE has addressed these needs through development of a comprehensive program plan, an LTS and M plan for each completed FUSRAP site, a records finding aid, and a public information web site. These developments help ensure that the current knowledge is preserved and passed on to future custodians and stakeholders. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
New Developments in the Technology Readiness Assessment Process in US DOE-EM - 13247
A Technology Readiness Assessment (TRA) is a systematic, metric-based process and accompanying report that evaluates the maturity of the technologies used in systems; it is designed to measure technology maturity using the Technology Readiness Level (TRL) scale pioneered by the National Aeronautics and Space Administration (NASA) in the 1980's. More recently, DoD has adopted and provided systematic guidance for performing TRAs and determining TRLs. In 2007 the GAO recommended that the DOE adopt the NASA/DoD methodology for evaluating technology maturity. Earlier, in 2006-2007, DOE-EM had conducted pilot TRAs on a number of projects at Hanford and Savannah River. In March 2008, DOE-EM issued a process guide, which established TRAs as an integral part of DOE-EM's Project Management Critical Decision Process. Since the development of its detailed TRA guidance in 2008, DOE-EM has continued to accumulate experience in the conduct of TRAs and the process for evaluating technology maturity. DOE has developed guidance on TRAs applicable department-wide. DOE-EM's experience with the TRA process, the evaluations that led to recently developed proposed revisions to the DOE-EM TRA/TMP Guide; the content of the proposed changes that incorporate the above lessons learned and insights are described. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Cementitious Barriers Partnership (CBP): Training and Release of CBP Toolbox Software, Version 1.0 - 13480
The Cementitious Barriers Partnership (CBP) Project is a multi-disciplinary, multi-institutional collaboration supported by the Office of Tank Waste Management within the Office of Environmental Management of U.S. Department of Energy (US DOE). The CBP program has developed a set of integrated tools (based on state-of-the-art models and leaching test methods) that improve understanding and predictions of the long-term hydraulic and chemical performance of cementitious barriers used in nuclear applications. Tools selected for and developed under this program are intended to evaluate and predict the behavior of cementitious barriers used in near-surface engineered waste disposal systems for periods of performance up to or longer than 100 years for operating facilities and longer than 1,000 years for waste management purposes. CBP software tools were made available to selected DOE Office of Environmental Management and field site users for training and evaluation based on a set of important degradation scenarios, including sulfate ingress/attack and carbonation of cementitious materials. The tools were presented at two-day training workshops held at U.S. National Institute of Standards and Technology (NIST), Savannah River, and Hanford included LeachXS{sup TM}/ORCHESTRA, STADIUM{sup R}, and a CBP-developed GoldSim Dashboard interface. Collectively, these components form the CBP Software Toolbox. The new U.S. Environmental Protection Agency leaching test methods based on the Leaching Environmental Assessment Framework (LEAF) were also presented. The CBP Dashboard uses a custom Dynamic-link library developed by CBP to couple to the LeachXS{sup TM}/ORCHESTRA and STADIUM{sup R} codes to simulate reactive transport and degradation in cementitious materials for selected performance assessment scenarios. The first day of the workshop introduced participants to the software components via presentation materials, and the second day included hands-on tutorial exercises followed by discussions of enhancements desired by participants. Tools were revised based on feedback obtained during the workshops held from April through June 2012. The resulting improved CBP Software Toolbox, including evaluation versions of and LeachXS{sup TM}/ORCHESTRA and STADIUM{sup R} has been made available to workshop and selected other participants for further assessment. Inquiries about future workshops and requests for access to the Toolbox software can be made via the CBP web site [1]. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Preliminary Analysis of Advanced Reactor Spent Nuclear Fuel Storage, Transportation, and Disposal
Due to increased interest in advanced reactor deployment and their associated potential new fuel cycles, the U.S. Department of Energy (DOE) Spent Fuel and Waste Science and Technology (SFWST) program has begun to evaluate the possible implications of long term management and final disposition of the spent nuclear fuel (SNF) generated. Safely managing and dispositioning this SNF, along with any other associated radioactive waste, is the primary focus of this initial preliminary assessment. This paper summarizes efforts to evaluate the characteristics and packaging options for three types of advanced reactor SNF forms: (1) tristructural isotropic (TRISO), (2) metallic, and (3) irradiated fuel salt presented in the report titled “Storage, Transportation, and Disposal of Advanced Reactor Spent Nuclear Fuel and High-Level Waste”. TRISO and metallic SNF and their associated waste streams were emphasized because of the near-term anticipated demonstrations of X-energy’s Xe-100 and TerraPower and GE Hitachi’s Natrium advanced reactors. Preliminary information on spent fuel salts discharged from molten-salt reactors (MSRs) was also examined to provide a baseline for future efforts. All calculations and assumptions were based on publicly available information. This paper identifies several different reactors that produce either TRISO or metallic SNF as well as a few of the reactor and fuel characteristics used for the assessments. Based on these characteristics, calculations were performed to determine the applicability of packaging SNF into existing canister designs. The evaluations included geometric (e.g., dimension, volume) and mass/weight considerations, known operational approaches and loading procedures, physical and chemical considerations/conditions for storage environments, and as-loaded radiation, thermal, and criticality analyses to identify constraints on storage, transportation, and disposal. Gaps in publicly available data pertaining to reactor operation and/or fuel composition provide increased uncertainty in some evaluations. Additionally, uncertainty in packaging and SNF management operations provide additional uncertainty. However, preliminary conclusions can still be assessed through this work and are presented in this paper.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Preliminary Analysis of Advanced Reactors Storage, Transportation, and Disposal
Based on the higher interest in Advanced Reactor (AR) deployment (e.g., ARDP[1]) for potential new fuel cycles, the Spent Fuel & Waste Science and Technology (SFWST) Program has begun to evaluate the possible implications of long-term management and final disposition of potential Advance Reactor spent nuclear fuels (SNF) that would be generated in potential advanced reactors. Safely managing and dispositioning the potential future AR SNF, and any other associated radioactive wastes, is the primary focus of this initial preliminary assessment of those. This paper summarizes the efforts by the Spent Fuel & Waste Science and Technology (SFWST) in evaluating characteristics and packaging options for advanced reactor spent nuclear fuel forms. The fuel forms were categorized into three types: (1) tristructural isotropic (TRISO), (2) metallic, and (3) fuel salt. This work emphasized TRISO and metallic SNF and waste streams because of the near-term anticipated operation of the Xe-100 and the Natrium reactors as advanced-reactor demonstrations. Preliminary information for the spent-fuel salts discharged from molten-salt reactors (MSRs) is also examined to provide a baseline for future efforts. All calculations and assumptions used publicly available information. The following characteristics are calculated or estimated for use in the preliminary assessments: SNF volume and mass, radiation/activity levels through time, thermal conditions through time, potential radionuclide source terms, chemical interactions and evolutions, disposal inventories, and waste-form lifetime. Using those characteristics, calculations to determine the applicability of existing canister designs were performed. These evaluations included geometric (e.g., dimension, volume) and mass/weight considerations, known operational approaches and loading procedures, physical and chemical considerations/conditions for storage environments, as-loaded radiation, thermal, and criticality analyses to identify constraints for storage, transportation, and disposal. The paper also includes a literature review and analysis on the storage, transportation, and disposal evaluations and experiences from reactors with similar fuel forms. Advanced-reactor vendors cite past experiences with Fort St. Vrain for TRISO and the Experimental Breeder Reactor II (EBR-II) for metallics that have major influences on fuel design. Finally, the paper includes preliminary concepts of operation for advanced-reactor SNF. This encompasses storage, transportation, potential treatment, and disposal activities from both a per-canister and systems-integration perspective.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Energy Efficiency and Conservation Block Grant (EECBG)- Better Buildings Neighborhood Program Final Report
The original BetterBuildings for Greensboro grant program included an outreach campaign to inform 100% of the Greensboro community about the benefits of reducing energy use; a plan to reduce energy consumption in at least 34% of the homes and 10% of the other buildings in the east Greensboro target area; and a plan to create and retain jobs in the energy conservation industry. Under the original program structure the City of Greensboro planned to partner with local and regional lenders to create a diversified portfolio of loan products to meet the needs of various income levels and building types. All participants would participate in the loan programs as a method of meeting the program’s 5 to1 private capital match/leverage requirements. In June 2011 the program was restructured to include partnerships with large commercial and multifamily projects, with these partners providing the greater portion of the required match/leverage. The geographic focus was revised to include reducing energy consumption across the entire City of Greensboro, targeting neighborhoods with high concentrations of low-moderate income households and aged housing stock. The community outreach component used a neighborhood-based approach to train community residents and volunteers to conduct door-to-door neighborhood sweeps; delivered high quality information on available program resources; helped residents to evaluate alternative energy efficiency measures and alternative financing sources; assisted with contractor selections and monitoring/evaluation of work; coordinated activities with BetterBuildings program partners; and collected data required by the Department of Energy. Additionally, HERO (Home Energy Response Officers) delivered intro packages (energy efficiency information and products) to thousands of households at the initial point of contact. A pilot program (Early Adopters) was offered from March 1, 2011 through June 30, 2011. The Early Adopters program was designed to offer immediate assistance to property owners ready and able to make their homes more energy efficient, by offering a rebate on their energy assessment and on the cost of upgrades installed. Eligible energy efficient upgrades were inclusive of basic level insulating and weather-stripping, HVAC system and water heater upgrades, to whole home upgrades that include the replacement of windows, doors and appliances. Renewable energy systems such as solar hot water systems were also eligible for the rebate program.
32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
Modernization of existing VVER-1000 surveillance programs
According to generally accepted world practice, evaluation of the reactor pressure vessel (RPV) material behavior during operation is carried out using tests of surveillance specimens. The main objective of the surveillance program consists in insurance of safe RPV operation during the design lifetime and lifetime-extension period. At present, the approaches of pressure vessels residual life validation based on the test results of their surveillance specimens have been developed and introduced in Russia and are under consideration in other countries where vodo-vodyanoi energetichesky reactors- (VVER-) 1000 are in operation. In this case, it is necessary to ensure leading irradiation of surveillance specimens (as compared to the pressure vessel wall) and to provide uniformly irradiated specimen groups for mechanical testing. Standard surveillance program of VVER-1000 has several significant shortcomings and does not meet these requirements. Taking into account program of lifetime extension of VVER-1000 operating in Russia, it is necessary to carry out upgrading of the VVER-1000 surveillance program. This paper studies the conditions of a surveillance specimen's irradiation and upgrading of existing sets to provide monitoring and prognosis of RPV material properties for extension of the reactor's lifetime up to 60 years or more. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
HTR-Proteus Pebble Bed Experimental Program Cores 9 & 10: Columnar Hexagonal Point-on-Point Packing with a 1:1 Moderator-to-Fuel Pebble Ratio
PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen critical configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
HTR-Proteus Pebble Bed Experimental Program Cores 5, 6, 7, & 8: Columnar Hexagonal Point-on-Point Packing with a 1:2 Moderator-to-Fuel Pebble Ratio
PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen critical configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
HTR-PROTEUS pebble bed experimental program cores 9 & 10: columnar hexagonal point-on-point packing with a 1:1 moderator-to-fuel pebble ratio
PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen critical configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
HTR-Proteus Pebble Bed Experimental Program Cores 5,6,7,&8: Columnar Hexagonal Point-on-Point Packing with a 1:2 Moderator-to-Fuel Pebble Ratio
PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen critical configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Automated design synthesis of robotic/human workcells for improved manufacturing system design in hazardous environments [Slides]
Manufacturing tasks that are deemed too hazardous for workers require the use of automation, robotics, and/or other remote handling tools. The associated hazards may be radiological or non-radiological, and based on the characteristics of the environment and processing, a design may necessitate robotic labor, human labor, or both. There are also other factors such as cost, ergonomics, maintenance, and efficiency that also effect task allocation and other design choices. Handling the tradeoffs of these factors can be complex, and lack of experience can be an issue when trying to determine if and what feasible automation/robotics options exist. To address this problem, we utilize common engineering design approaches adapted more for manufacturing system design in hazardous environments. We limit our scope to the conceptual and embodiment design stages, specifically a computational algorithm for concept generation and early design evaluation. In regard to concept generation, we first develop the functional model or function structure for the process, using the common 'verb-noun' format for describing function. A common language or functional basis for manufacturing was developed and utilized to formalize function descriptions and guide rules for function decomposition. Potential components for embodiment are also grouped in terms of this functional language and are stored in a database. The properties of each component are given as quantitative and qualitative criteria. Operators are also rated for task-relevant criteria which are used to address task compatibility. Through the gathering of process requirements/constraints, construction of the component database, and development of the manufacturing basis and rule set, design knowledge is stored and available for computer use. Thus, once the higher level process functions are defined, the computer can automate the synthesis of new design concepts through alternating steps of embodiment and function structure updates/decomposition. In the process, criteria guide function allocation of components/operators and help ensure compatibility and feasibility. Through multiple function assignment options and varied function structures, multiple design concepts are created. All of the generated designs are then evaluated based on a number of relevant evaluation criteria: cost, dose, ergonomics, hazards, efficiency, etc. These criteria are computed using physical properties/parameters of each system based on the qualities an engineer would use to make evaluations. Nuclear processes such as oxide conversion and electrorefining are utilized to aid algorithm development and provide test cases for the completed program. Through our approach, we capture design knowledge related to manufacturing and other operations in hazardous environments to enable a computational program to automatically generate and evaluate system design concepts.
42 ENGINEERING↗
Improving the Advancement of Women in Computer and Computational Science Research with the CRA-W Career Mentoring Workshops (Final Report)
The mission of the Computing Research Association’s Committee on Widening Participation in Computing Research (CRA-WP) is to widen the participation and improve the access, opportunities, and positive experiences of individuals from groups underrepresented in computing research and education. CRA-WP programs serve this overarching goal at all career stages; in addition, CRA-WP, through the formation of the CRA Center for Evaluating the Research Pipeline (CERP), has developed a methodology for thoroughly evaluating the success of its programs by comparing a nationwide sample of students, researchers, and faculty (non-participants) to program participants. Achieving these objectives requires that an increasing number of individuals from populations underrepresented in computing start and progress to the next stage while understanding and supporting the myriad computing pathways. CRA-WP offers programs for participants from undergraduate to senior professional levels. Different career stages need different types of interventions, and the goal of all Alliance program activities can be described within the unifying framework of Social Cognitive Career Theory, which finds that interest in and choice of a particular career path will be increased by interventions that improve one or more of the following: (1) outcome expectations (understanding and valuing the rewards of a particular outcome), (2) self-efficacy (a belief that one can successfully achieve an outcome), and (3) social supports that help one persist and overcome obstacles.
97 MATHEMATICS AND COMPUTING↗
Light Water Reactor Sustainability Program, U.S. Efforts in Support of Examinations at Fukushima Daiichi-2017 Evaluations
Although the accident signatures from each unit at the Fukushima Daiichi Nuclear Power Station (NPS) [Daiichi] differ, much is not known about the end-state of core materials within these units. Some of this uncertainty can be attributed to a lack of information related to cooling system operation and cooling water injection. There is also uncertainty in our understanding of phenomena affecting: a) in-vessel core damage progression during severe accidents in boiling water reactors (BWRs), and b) accident progression after vessel failure (ex-vessel progression) for BWRs and Pressurized Water Reactors (PWRs). These uncertainties arise due to limited full scale prototypic data. Similar to what occurred after the accident at Three Mile Island Unit 2, these Daiichi units offer the international community a means to reduce such uncertainties by obtaining prototypic data from multiple full-scale BWR severe accidents. Information obtained from Daiichi is required to inform Decontamination and Decommissioning activities, improving the ability of the Tokyo Electric Power Company Holdings, Incorporated (TEPCO Holdings) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This document, which has been updated to include FY2017 information, summarizes results from U.S. efforts to use information obtained by TEPCO Holdings to enhance the safety of existing and future nuclear power plant designs. This effort, which was initiated in 2014 by the Reactor Safety Technologies Pathway of the Department of Energy Office of Nuclear Energy Light Water Reactor (LWR) Sustainability Program, consists of a group of U.S. experts in LWR safety and plant operations that have identified examination needs and are evaluating TEPCO Holdings information from Daiichi that address these needs. Each year, annual reports include examples demonstrating that significant safety insights are being obtained in the areas of component performance, fission product release and transport, debris end-state location, and combustible gas generation and transport. In addition to reducing uncertainties related to severe accident modeling progression, these insights are being used to update guidance for severe accident prevention, mitigation, and emergency planning. Furthermore, reduced uncertainties in modeling the events at Daiichi will improve the realism of reactor safety evaluations and inform future D&D activities by improving the capability for characterizing potential hazards to workers involved with cleanup activities. Highlights in this FY2017 report include new insights with respect to the forces required to produce the observed Daiichi Unit 1 (1F1) shield plug endstate, the observed leakage from 1F1 components, and the amount of combustible gas generation required to produce the observed explosions in Daiichi Units 3 and 4 (1F3 and 1F4). This report contains an appendix with a list of examination needs that was updated after U.S. experts reviewed recently obtained information from examinations at Daiichi. Additional details for higher priority, near-term, examination activities are also provided. This report also includes an appendix with a description of an updated website that has been reformatted to better assist U.S. experts by providing information in an archived retrievable location, as well as an appendix summarizing U.S. Forensics activities to host the TMI-2 Knowledge Transfer and Relevance to Fukushima Meeting that was held in Idaho Falls, ID, on October 10-14, 2016.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗