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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

ENRESA's Spent Nuclear Fuel and Special Waste Database - 18115

Empresa Nacional de Residuos Radiactivos S.A. (ENRESA, by its initials in Spanish) Spent Nuclear Fuel (SNF) and Special Waste (SW) database development was initiated in a common effort between ENRESA and UNESA (the Spanish Association of the Electricity Industry) as a consequence of new regulatory requirements in the frame of radioactive waste management plans. Although these requirements were only mandatory for operators of nuclear power plants, ENRESA deemed necessary to have a centralized database which collected the data from the plant operators. The most relevant features of this database are: - The information collected covers not only the general characteristics of the SNF and SW but also the operation details needed for their characterization. - An application, named Gestion de Elementos Combustibles y Residuos Especiales (GECYRE, by its initials in Spanish) has been implemented for easing the queries and the generation of predefined reports. - An agile data transfer scheme, based on XML files, is used for an effective communication between ENRESA and the operators. The database is currently in production phase since 2014. During this phase it has demonstrated its effectiveness as an information repository to address the SNF and SW interim storage projects in Spain foreseen in the near future. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2010

Methods for Processing ENDF/B-VII with NJOY

The NJOY Nuclear Data Processing System is widely used to convert evaluations in the Evaluated Nuclear Data Files (ENDF) format into forms useful for practical applications such as fission and fusion reactor analysis, stockpile stewardship calculations, criticality safety, radiation shielding, nuclear waste management, nuclear medicine procedures, and more. This paper provides a description of the system's capabilities, summary descriptions of the methods used, and information on how to use the code to process the modern evaluated nuclear data files from ENDF/B-VII. It begins with the generation of pointwise libraries, including reaction and resonance reconstruction, Doppler broadening, radiation heating and damage, thermal scattering data, unresolved resonance data, and gas production. It then reviews the production of libraries for the continuous-energy Monte Carlo code MCNP, multigroup neutron, photon, and particle cross sections and matrices, and photon interaction data. The generation of uncertainty information for ENDF data is discussed, including new capabilities for calculating covariances of resonance data, angular distributions, energy distributions, and radioactive nuclide production. NJOY's ability to prepare thermal scattering data evaluations for bound moderators (which was used during the preparation of the ENDF/B-VII library) is described. The strong plotting capabilities of NJOY are summarized. Many examples of black and white and color Postscript plots are included throughout the paper. The capabilities of NJOY to output multigroup data in several different formats to suit various applications is reviewed. Finally, a section is included that summarizes the history of the development of the NJOY system over the last 37 years.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

OECD NEA radioactive waste repository metadata management (RepMet) initiative (2014-2018) - 15614

The OECD/NEA Radioactive Waste Repository Metadata Management (RepMet) initiative aims to bring about a better understanding of the identification and administration of metadata - a key aspect of data management - to support national programmes in managing their radioactive waste repository data, information and records in a way that is both harmonised internationally and suitable for long-term management and use. At this point in the project, the main focus lies on 'waste packages at their storage sites ready for geological disposal' but excluding spent fuel and high level waste (HLW). Later work will deal with metadata collected on spent fuel and HLW as well as with different phases of repository development: design and construction, and operation. In order to develop metadata, RepMet members have provided lists of data items they currently collect and manage for packaged waste. The lists, albeit it non comprehensive, suggest that the number and type of data items varies importantly from organisation to organisation. At term, RepMet plans to generate a comprehensive list of such data items and their metadata. In parallel, RepMet is developing a document which examines a selected number of existing metadata standards. Because of the specialist nature of radioactive waste management, including waste processing facilities, disposal facilities and waste packages, existing metadata standards do not seem to be transferable as defined. Rather, RepMet will explore such standards to identify relevant principles for possible adaptation to the radioactive waste management community. (authors)

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

V and V of the HTR Code Package (HCP) as a comprehensive HTR steady state and transient safety analysis framework - 18664

The HTR code package (HCP) allows for the simulation of several safety-related aspects of a High Temperature Reactor core in a highly integrated manner. HCP currently couples the thermo-fluid dynamics and time dependent neutronics code MGT, the spectrum code TRISHA, the burn up code TNT, the fuel management code SHUFLE and the fission product release code STACY. During its development, state-of-the-art programming techniques and standards are applied. The cross sections in HCP are generated by a 0D- or 1D solver featuring an innovative approach to treat the double heterogeneity of pebble fuel. Also major improvements have been made to optimize the nuclear data library based on ENDF/B-VII. Two advanced fuel management models are provided by the code module SHUFLE that go far beyond the capabilities of existing system codes like VSOP. The source term analysis code module STACY is coupled to HCP providing release rate calculations with a high spatial resolution making use of the nuclide densities provided by TNT. An outstanding new feature of HCP is the possibility to simulate long-term operation scenarios based on OTTO or MEDUL fuel shuffling schemes as well as selected transients in one integrated code package. Even alternating steady state/transient/steady state simulations are possible. So with HCP different fuel strategies and their influence on various kinds of accidents can be examined with one consistent reactor model. This paper provides an overview of the development status of the HCP and reports about selected benchmark results. It is demonstrated that the new system code HCP is capable to replace existing stand-alone codes like VSOP, TINTE/MGT, FRESCO or PANAMA while introducing new features, which so far to our knowledge were not available in the field of HTR safety research. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Review of the Nuclear Energy Agency (NEA) Ancillary Thermodynamic Database (TDB) Volume (DRAFT REV. 0)

The Nuclear Energy Agency (NEA) Ancillary data volume comprises thermodynamic data of mineral and aqueous species that, in addition to Auxiliary Data (as referred to in previous NEA thermodynamic data volumes), is necessary to calculations of chemical interactions relevant to radioactive waste management and nuclear energy. This SAND report is a review of the NEA Ancillary data critical reviews volume of thermodynamic data parameters. The review given in this report mainly involves data comparison with other thermodynamic data assessments, analysis of thermodynamic parameters, and examination of data sources. Only new and updated data parameters were considered in this review. Overall, no major inconsistencies or errors were found as allowed by the comparisons conducted in this review. Some remarks were noted, for example, on the consideration of relevant studies and/or comparisons on the analysis and retrieval of thermodynamic data parameters not cited in the respective sections.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Data Driven Analysis for Modernization Program Management

NNSA is responsible for managing national nuclear security missions: ensuring a safe, secure, and reliable nuclear deterrent; supplying nuclear fuel to the Navy; and supporting the nation’s nuclear nonproliferation efforts. However, over half of NNSA’s facilities are more than 40 years old, and roughly one-third date back to the Manhattan Project. To execute its critical nuclear security missions, NNSA is making large investments to modernize its nuclear production capabilities. This ramp up represents NNSA’s largest modernization effort since the Cold War. Given the scale of these efforts, NNSA’s Office of Secondary Stage Production Modernization has implemented data driven techniques to prioritize investments and inform strategic decision making. NNSA, with support from its site managing contractors, has developed and implemented an integrated schedule and risk management system to address the issues and limitations with the traditional approach. The multi-year integrated schedules are key to identifying program linkages and managing large portfolios comprised of many different projects and efforts. In conjunction with the integrated schedule, a new program risk management system has also been developed and implemented, which manages program risks and opportunities, along with specific mitigation strategies to reduce or eliminate the risks per timelines that are tracked in the integrated schedule. This paper is supplemented with a follow-on presentation on effective management of program material and throughput modeling.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

International Cooperation in Nuclear Data Evaluation

The OECD Nuclear Energy Agency (NEA) is organising a co-operation between the major nuclear data evaluation projects in the world. The co-operation involves the ENDF, JEFF, and JENDL projects, and, owing to the collaboration with the International Atomic Energy Agency (IAEA), also the Russian BROND and the Chinese CENDL projects. The Working Party on international nuclear data Evaluation Cooperation (WPEC), comprised of about 20 core members, manages this co-operation and meets annually to discuss progress in each evaluation project and also related experimental activities. The WPEC assesses common needs for nuclear data improvements and these needs are then addressed by initiating joint evaluation efforts. The work is performed in specially established subgroups, consisting of experts from the participating evaluation projects. The outcome of these subgroups is published in reports, issued by the NEA. Current WPEC activities comprise for example a number of studies related to nuclear data uncertainties, including a review of methods for the combined use of integral experiments and covariance data, as well as evaluations of some of the major actinides, such as {sup 235}U and {sup 239}Pu. This paper gives an overview of current and planned activities within the WPEC.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2012

Uncertainty quantification for accident management using ACE surrogates

The alternating conditional expectation (ACE) regression method is used to generate RELAP5 surrogates which are then used to determine the distribution of the peak clad temperature (PCT) during the loss of feedwater accident coupled with a subsequent initiation of the feed and bleed (F and B) operation in the Zion-1 nuclear power plant. The construction of the surrogates assumes conditional independence relations among key reactor parameters. The choice of parameters to model is based on the macroscopic balance statements governing the behavior of the reactor. The peak clad temperature is calculated based on the independent variables that are known to be important in determining the success of the F and B operation. The relationship between these independent variables and the plant parameters such as coolant pressure and temperature is represented by surrogates that are constructed based on 45 RELAP5 cases. The time-dependent PCT for different values of F and B parameters is calculated by sampling the independent variables from their probability distributions and propagating the information through two layers of surrogates. The results of our analysis show that the ACE surrogates are able to satisfactorily reproduce the behavior of the plant parameters even though a quasi-static assumption is primarily used in their construction. The PCT is found to be lower in cases where the F and B operation is initiated, compared to the case without F and B, regardless of the F and B parameters used. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Light Water Reactor Sustainability Program: Reactor Safety Technologies Pathway Technical Program Plan

In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safety initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

SORTIE DB - Source term integrated environmental management disposal facility database - 15249

The Environmental Management Disposal Facility (EMDF) in Oak Ridge, Tennessee, is a proposed U.S. Department of Energy low-level Comprehensive Environmental Response Compensation and Liability Act (CERCLA) waste disposal facility with an operational start in the early 2020's. As a DOE low-level waste disposal facility, it is subject to DOE Order 435.1, requiring a composite analysis and possibly a performance assessment. The Source Term Integrated EMDF Data Base - SORTIE DB - provides source term estimates used in these assessments. An Oak Ridge Office of Environmental Management (OREM) Federal and Contractor team developed this state-of-the-art descriptive analytics and visualization schema. DOE OREM senior management has recognized complementary beneficial uses of SORTIE DB in decisions for accelerating clean-up progress, dealing with conflicting sequencing and budget priorities, and enhancing stakeholder and regulatory interfaces. SORTIE DB is designated as an OREM planning tool. SORTIE DB contains estimated forecast waste volumes and masses for over 800 Decontamination and Decommissioning (D and D) and Remedial Action (RA) sequenced baseline packages over the next 25+ years. Available characterization data, radionuclide identification and concentrations, and data from safety basis and emergency management documents are incorporated into SORTIE DB. (authors)

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

Choosing Your Nuclear Fuel Cycle: A Life Cycle Assessment Perspective - 16425

Nuclear waste management has always been a much debated topic and a key driver for decisions within the nuclear industry. A standardised global approach has not yet been developed and at present many countries are re-considering their position. In the UK the reprocessing of Spent Nuclear Fuel is projected to stop, but no clear indication about the future approach has been given. This contribution sets out the early stages of an approach to compare the possible alternatives based on Life Cycle Assessment (LCA), which is widely used in other sectors to assess the overall environmental burdens of a product or a service across the whole life cycle - i.e. from cradle to grave. The LCA may serve to support decision-making processes within the nuclear industry and, provided that it is used in an open and transparent way, to improve public attitudes towards nuclear energy. To date few LCA studies within the nuclear industry have been carried out, the main issue being the lack of a standardised methodology to evaluate the impacts of radionuclides. The present study proposes a new framework for this purpose, and demonstrates its application in an LCA study of the current UK approach to the management of spent nuclear fuel based on specific data from plants at Sellafield. The final results of the study will show the overall environmental footprint of the reprocessing of spent nuclear fuel, in terms of both radiological and non-radiological impacts. A 'hot spot' analysis will also be performed to highlight the critical processing steps. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2010

NGNP Data Management and Analysis System Analysis and Web Delivery Capabilities

Projects for the Very High Temperature Reactor Technology Development Office provide data in support of Nuclear Regulatory Commission licensing of the very high temperature reactor. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high-temperature and high-fluence environments. In addition, thermal-hydraulic experiments are conducted to validate codes used to assess reactor safety. The Very High Temperature Reactor Technology Development Office has established the NGNP Data Management and Analysis System (NDMAS) at the Idaho National Laboratory to ensure that very high temperature reactor data are (1) qualified for use, (2) stored in a readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the third NDMAS objective. It describes capabilities for displaying the data in meaningful ways and for data analysis to identify useful relationships among the measured quantities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Streamlining analysis capabilities for SNF management - 15490

A comprehensive, integrated data and analysis tool-the Used Nuclear Fuel-Storage, Transportation and Disposal Analysis Resource and Data System (UNF-ST and DARDS) -is being developed for the US DOE Office of Nuclear Energy (DOE-NE) Nuclear Fuels Storage and Transportation Planning Project (NFST). The overarching goal of UNF-ST and DARDS is to provide a comprehensive controlled source of technical data integrated with key analysis capabilities to characterize inputs to the overall US waste management system from reactor power production through ultimate disposition. UNF-ST and DARDS seamlessly integrates a unified SNF system relational database and key analysis capabilities to simplify and automate the performance of accurate and efficient SNF related analyses to support numerous DOE waste management and fuel cycle activities. Hence, UNF-ST and DARDS enables decision making relative to design, safety, and licensing of SNF systems and facilities by providing the best information available. Automated analysis sequences have been deployed to characterize the discharge inventory and perform criticality, thermal, and dose analyses using time-dependent data. UNF-ST and DARDS provides cask-specific as-loaded safety analysis including criticality, thermal and dose evaluations, as well as results visualization. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Current Reactor Physics Benchmark Activities at the Idaho National Laboratory

The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) were established to preserve integral reactor physics and criticality experiment data for present and future research. These valuable assets provide the basis for recording, developing, and validating our integral nuclear data, and experimental and computational methods. These projects are managed through the Idaho National Laboratory (INL) and the Organisation for Economic Co-operation and Development Nuclear Energy Agency (OECD-NEA). Staff and students at the Department of Energy - Idaho (DOE-ID) and INL are engaged in the development of benchmarks to support ongoing research activities. These benchmarks include reactors or assemblies that support Next Generation Nuclear Plant (NGNP) research, space nuclear Fission Surface Power System (FSPS) design validation, and currently operational facilities in Southeastern Idaho.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
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)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Data Analytics for Concrete Structural Health Monitoring in Nuclear Power Plants

Assessment and management of aging concrete structures in nuclear power plants require a more systematic approach than simple reliance on existing code margins of safety. Structural health monitoring of concrete structures aims to understand the current health condition of a structure based on heterogeneous measurements to produce high-confidence actionable information regarding structural integrity that supports operational and maintenance decisions. This ongoing research project is seeking to develop a probabilistic framework for health diagnosis and prognosis of aging concrete structures in a nuclear power plant that is subjected to physical, chemical, environment, and mechanical degradation. In this paper, experimental set up at Vanderbilt University’s Laboratory for Systems Integrity and Reliability is used to research effective combinations of full field techniques, which include infrared thermography, digital image correlation, and ultrasonic measurement. The full-field imaging data collected were processed using contour mapping technique to identify holes in a concrete slab. The techniques discussed and presented in this paper allow diagnosis of damage. The framework will enable plant operators to make risk informed decisions on the struc-tural integrity, remaining useful life, and performance of concrete structures.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Final Report for the Workshop for Applied Nuclear Data Activities

The Workshop for Applied Nuclear Data Activities (WANDA) was held at the Elliot School of International Affairs at George Washington University from 22 to 25 January 2019. The purpose of WANDA was to bring subject matter experts from the national laboratories, universities and industry together with government program managers and their advisors to develop collaborative plans of action (e.g., roadmaps ) to address outstanding issues in nuclear data that effect applications ranging from nuclear energy to national security and nonproliferation to isotope production. WANDA was the latest in a series of collaborative, cross-programmatic workshops focused on nuclear data needs for applications that started in 2015 with the Nuclear Data Needs and Capabilities for Applications Workshop (NDNCA) and continuing with the Nuclear Data Roadmapping Enhancement Workshop (NDREW) in 2018 focused on nuclear data needs for applications. NDREW and WANDA played an important role in providing guidance for the recently created Nuclear Data Interagency Working Group (NDIAWG) comprised of program managers from a panoply government agencies. The NDIAWG issued two Funding Opportunity Announcements (FOA) in FY2017 and FY2018 which are supporting collaborative efforts to address high-priority nuclear data needs.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2012

Uranium Enrichment Standards of the Y-12 Nuclear Detection and Sensor Testing Center

The Y-12 National Security Complex has recently fabricated and characterized a new series of metallic uranium standards for use in the Nuclear Detection and Sensor Testing Center (NDSTC). Ten uranium metal disks with enrichments varying from 0.2 to 93.2% {sup 235}U were designed to provide researchers access to a wide variety of measurement scenarios in a single testing venue. Special care was taken in the selection of the enrichments in order to closely bracket the definitions of reactor fuel at 4% {sup 235}U and that of highly enriched uranium (HEU) at 20% {sup 235}U. Each standard is well characterized using analytical chemistry as well as a series of gamma-ray spectrometry measurements. Gamma-ray spectra of these standards are being archived in a reference library for use by customers of the NDSTC. A software database tool has been created that allows for easier access and comparison of various spectra. Information provided through the database includes: raw count data (including background spectra), regions of interest (ROIs), and full width half maximum calculations. Input is being sought from the user community on future needs including enhancements to the spectral database and additional Uranium standards, shielding configurations and detector types. A related presentation are planned for the INMM 53rd Annual Meeting (Hull, et al.), which describe new uranium chemical compound standards and testing opportunities at Y-12 Nuclear Detection and Sensor Testing Center (NDSTC).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗