Sample records for nuclear plant ngnp

  1. The Next Generation Nuclear Plant (NGNP) Project

    SciTech Connect (OSTI)

    F. H. Southworth; P. E. MacDonald

    2003-11-01T23:59:59.000Z

    The Next Generation Nuclear Power (NGNP) Project will demonstrate emissions-free nuclearassisted electricity and hydrogen production by 2015. The NGNP reactor will be a helium-cooled, graphite moderated, thermal neutron spectrum reactor with a design goal outlet temperature of 1000 C or higher. The reactor thermal power and core configuration will be designed to assure passive decay heat removal without fuel damage during hypothetical accidents. The fuel cycle will be a once-through very high burnup low-enriched uranium fuel cycle. This paper provides a description of the project to build the NGNP at the Idaho National Engineering and Environmental Laboratory (INEEL). The NGNP Project includes an overall reactor design activity and four major supporting activities: materials selection and qualification, NRC licensing and regulatory support, fuel development and qualification, and the hydrogen production plant. Each of these activities is discussed in the paper. All the reactor design and construction activities will be managed under the DOE’s project management system as outlined in DOE Order 413.3. The key elements of the overall project management system discussed in this paper include the client and project management organization relationship, critical decisions (CDs), acquisition strategy, and the project logic and timeline. The major activities associated with the materials program include development of a plan for managing the selection and qualification of all component materials required for the NGNP; identification of specific materials alternatives for each system component; evaluation of the needed testing, code work, and analysis required to qualify each identified material; preliminary selection of component materials; irradiation of needed sample materials; physical, mechanical, and chemical testing of unirradiated and irradiated materials; and documentation of final materials selections. The NGNP will be licensed by the NRC under 10 CFR 50 or 10 CFR 52, for the purpose of demonstrating the suitability of high-temperature gas-cooled reactors for commercial electric power and hydrogen production. Products that will support the licensing of the NGNP include the environmental impact statement, the preliminary safety analysis report, the NRC construction permit, the final safety analysis report, and the NRC operating license. The fuel development and qualification program consists of five elements: development of improved fuel manufacturing technologies, fuel and materials irradiations, safety testing and post-irradiation examinations, fuel performance modeling, and fission product transport and source term modeling. Two basic approaches will be explored for using the heat from the high-temperature helium coolant to produce hydrogen. The first technology of interest is the thermochemical splitting of water into hydrogen and oxygen. The most promising processes for thermochemical splitting of water are sulfur-based and include the sulfur-iodine, hybrid sulfur-electrolysis, and sulfur-bromine processes. The second technology of interest is thermally assisted electrolysis of water. The efficiency of this process can be substantially improved by heating the water to high-temperature steam before applying electrolysis.

  2. Nuclear Safeguards Infrastructure Required for the Next Generation Nuclear Plant (NGNP)

    SciTech Connect (OSTI)

    Dr. Mark Schanfein; Philip Casey Durst

    2012-07-01T23:59:59.000Z

    The Next Generation Nuclear Plant (NGNP) is a Very High Temperature Gas-Cooled Reactor (VHTR) to be constructed near Idaho Falls, Idaho The NGNP is intrinsically safer than current reactors and is planned for startup ca. 2021 Safety is more prominent in the minds of the Public and Governing Officials following the nuclear reactor meltdown accidents in Fukushima, Japan The authors propose that the NGNP should be designed with International (IAEA) Safeguards in mind to support export to Non-Nuclear-Weapons States There are two variants of the NGNP design; one using integral Prismatic-shaped fuel assemblies in a fixed core; and one using recirculating fuel balls (or Pebbles) The following presents the infrastructure required to safeguard the NGNP This infrastructure is required to safeguard the Prismatic and Pebble-fueled NGNP (and other HTGR/VHTR) The infrastructure is based on current Safeguards Requirements and Practices implemented by the International Atomic Energy Agency (IAEA) for similar reactors The authors of this presentation have worked for decades in the area of International Nuclear Safeguards and are recognized experts in this field Presentation for INMM conference in July 2012.

  3. Next Generation Nuclear Plant (NGNP) Prismatic HTGR Conceptual Design Project - Final Technical Report

    SciTech Connect (OSTI)

    Saurwein, John

    2011-07-15T23:59:59.000Z

    This report is the Final Technical Report for the Next Generation Nuclear Plant (NGNP) Prismatic HTGR Conceptual Design Project conducted by a team led by General Atomics under DOE Award DE-NE0000245. The primary overall objective of the project was to develop and document a conceptual design for the Steam Cycle Modular Helium Reactor (SC-MHR), which is the reactor concept proposed by General Atomics for the NGNP Demonstration Plant. The report summarizes the project activities over the entire funding period, compares the accomplishments with the goals and objectives of the project, and discusses the benefits of the work. The report provides complete listings of the products developed under the award and the key documents delivered to the DOE.

  4. Engineering Design Elements of a Two-Phase Thermosyphon to Trannsfer NGNP Nuclear Thermal Energy to a Hydrogen Plant

    SciTech Connect (OSTI)

    Piyush Sabharwal

    2009-07-01T23:59:59.000Z

    Two hydrogen production processes, both powered by a Next Generation Nuclear Plant (NGNP), are currently under investigation at Idaho National Laboratory. The first is high-temperature steam electrolysis, which uses both heat and electricity; the second is thermo-chemical production through the sulfur iodine process primarily using heat. Both processes require a high temperature (>850°C) for enhanced efficiency; temperatures indicative of the NGNP. Safety and licensing mandates prudently dictate that the NGNP and the hydrogen production facility be physically isolated, perhaps requiring separation of over 100 m.

  5. Site Selection & Characterization Status Report for Next Generation Nuclear Plant (NGNP)

    SciTech Connect (OSTI)

    Mark Holbrook

    2007-09-01T23:59:59.000Z

    In the near future, the US Department of Energy (DOE) will need to make important decisions regarding design and construction of the Next Generation Nuclear Plant (NGNP). One part of making these decisions is considering the potential environmental impacts that this facility may have, if constructed here at the Idaho National Laboratory (INL). The National Environmental Policy Act (NEPA) of 1969 provides DOE decision makers with a process to systematically consider potential environmental consequences of agency decisions. In addition, the Energy Policy Act of 2005 (Title VI, Subtitel C, Section 644) states that the 'Nuclear Regulatory Commission (NRC) shall have licensing and regulatory authority for any reactor authorized under this subtitle.' This stipulates that the NRC will license the NGNP for operation. The NRC NEPA Regulations (10 CFR Part 51) require tha thte NRC prepare an Environmental Impact Statement (EIS) for a permit to construct a nuclear power plant. The applicant is required to submit an Environmental report (ER) to aid the NRC in complying with NEPA.

  6. NEXT GENERATION NUCLEAR PLANT NGNP Technology Development Roadmapping

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    ISR Inner Side Reflector Kc Fracture Toughness kg Kilogram K-T Kepner-Tregoe KTA German nuclear technical committee kW Kilowatt LANL Los Alamos National Laboratory LBE Licensing...

  7. Summary of Bounding Requirements for the NGNP Demonstration Plant F&ORs

    SciTech Connect (OSTI)

    Not Listed

    2008-06-01T23:59:59.000Z

    This report documents bounding functional and operating requirements (F&ORs)for the Next Generation Nuclear Plant (NGNP) Project to support selection of the nuclear system design and specification of the operating conditions and configuration of NGNP once the nuclear system design is selected. These requirements supplement the detailed F&ORs for NGNP developed in the FY07 NGNP Pre-conceptual design work.

  8. NGNP: High Temperature Gas-Cooled Reactor Key Definitions, Plant Capabilities, and Assumptions

    SciTech Connect (OSTI)

    Phillip Mills

    2012-02-01T23:59:59.000Z

    This document is intended to provide a Next Generation Nuclear Plant (NGNP) Project tool in which to collect and identify key definitions, plant capabilities, and inputs and assumptions to be used in ongoing efforts related to the licensing and deployment of a high temperature gas-cooled reactor (HTGR). These definitions, capabilities, and assumptions are extracted from a number of sources, including NGNP Project documents such as licensing related white papers [References 1-11] and previously issued requirement documents [References 13-15]. Also included is information agreed upon by the NGNP Regulatory Affairs group's Licensing Working Group and Configuration Council. The NGNP Project approach to licensing an HTGR plant via a combined license (COL) is defined within the referenced white papers and reference [12], and is not duplicated here.

  9. NGNP: High Temperature Gas-Cooled Reactor Key Definitions, Plant Capabilities, and Assumptions

    SciTech Connect (OSTI)

    Wayne Moe

    2013-05-01T23:59:59.000Z

    This document provides key definitions, plant capabilities, and inputs and assumptions related to the Next Generation Nuclear Plant to be used in ongoing efforts related to the licensing and deployment of a high temperature gas-cooled reactor. These definitions, capabilities, and assumptions were extracted from a number of NGNP Project sources such as licensing related white papers, previously issued requirement documents, and preapplication interactions with the Nuclear Regulatory Commission (NRC).

  10. New Generation Nuclear Plant -- High Level Functions and Requirements

    SciTech Connect (OSTI)

    J. M. Ryskamp; E. J. Gorski; E. A. Harvego; S. T. Khericha; G. A. Beitel

    2003-09-01T23:59:59.000Z

    This functions and requirements (F&R) document was prepared for the Next Generation Nuclear Plant (NGNP) Project. The highest-level functions and requirements for the NGNP preconceptual design are identified in this document, which establishes performance definitions for what the NGNP will achieve. NGNP designs will be developed based on these requirements by commercial vendor(s).

  11. Next Generation Nuclear Plant Phenomena

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the U.S. Department of Energy (DOE) to develop jointly a licensing strategy for the Next Generation Nuclear plant (NGNP), a very high temperature gas-cooled reactor (VHTR) for...

  12. NGNP Research and Development Status

    SciTech Connect (OSTI)

    David A. Petti

    2010-08-01T23:59:59.000Z

    At the inception of the Next Generation Nuclear Plant (NGNP) project, experts from the Department of Energy (DOE) national laboratories, gas reactor vendors, and universities collaborated to establish technology research and development (R&D) roadmaps. These roadmaps outlined the testing and computational development activities needed to qualify the materials and validate the modeling and simulation tools to be used in the design and safe operation of the NGNP, a helium-cooled, high temperature gas reactor (HTGR).

  13. Microsoft Word - NGNP-NHS-HTS-RPT-M-00004-R0-091809 .doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    List of DDN Revisions for the 750C-800C IHX ... 142 Next Generation Nuclear Plant: NGNP-NHS-HTS-RPT-M-00004 Intermediate Heat Exchanger...

  14. NEXT GENERATION NUCLEAR PLANT PROJECT IMPLEMENTATION STRATEGY

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    NEXT GENERATION NUCLEAR PLANT PROJECT IMPLEMENTATION STRATEGY Presented by NGNP Industry Alliance November 30, 2009 I In nd du us st tr ry y A Al ll li ia an nc ce e Clean,...

  15. NGNP Engineering Status

    SciTech Connect (OSTI)

    John Collins

    2010-08-01T23:59:59.000Z

    The objectives of Phase 1 Engineering and Design scope are to: 1) complete the initial design activities for a prototype nuclear reactor and plant that is capable of co-generating electricity, hydrogen, and process heat; 2) identify technological aspects of the NGNP that need further advancement by research and development activities; and 3) provide engineering support to the early licensing process, including technical input to white papers and developing the basis for future safety analyses.

  16. New Materials for NGNP/Gen IV

    SciTech Connect (OSTI)

    Robert W. Swindeman; Douglas L. Marriott

    2009-12-18T23:59:59.000Z

    The bounding conditions were briefly summarized for the Next Generation Nuclear Plant (NGNP) that is the leading candidate in the Department of Energy Generation IV reactor program. Metallic materials essential to the successful development and proof of concept for the NGNP were identified. The literature bearing on the materials technology for high-temperature gas-cooled reactors was reviewed with emphasis on the needs identified for the NGNP. Several materials were identified for a more thorough study of their databases and behavioral features relative to the requirements ASME Boiler and Pressure Vessel Code, Section III, Division 1, Subsection NH.

  17. Next Generation Nuclear Plant Materials Selection and Qualification Program Plan

    SciTech Connect (OSTI)

    R. Doug Hamelin; G. O. Hayner

    2004-11-01T23:59:59.000Z

    The U.S. Department of Energy (DOE) has selected the Very High Temperature Reactor (VHTR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production without greenhouse gas emissions. The reactor design is a graphite-moderated, helium-cooled, prismatic or pebble bed thermal neutron spectrum reactor with an average reactor outlet temperature of at least 1000 C. The NGNP will use very high burn up, lowenriched uranium, TRISO-Coated fuel in a once-through fuel cycle. The design service life of the NGNP is 60 years.

  18. NGNP-HPS SHAW-HPA-000 NGNP Hydrogen Plant Alternatives Study

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    of Energy DRL Design Readiness Level EDI Electro-Deionization FUS Feed and Utility Systems HAD Hydroiodic Acid Decomposition HHV High Heating Value HPAS Hydrogen Plant...

  19. NGNP Infrastructure Readiness Assessment: Consolidation Report

    SciTech Connect (OSTI)

    Brian K Castle

    2011-02-01T23:59:59.000Z

    The Next Generation Nuclear Plant (NGNP) project supports the development, demonstration, and deployment of high temperature gas-cooled reactors (HTGRs). The NGNP project is being reviewed by the Nuclear Energy Advisory Council (NEAC) to provide input to the DOE, who will make a recommendation to the Secretary of Energy, whether or not to continue with Phase 2 of the NGNP project. The NEAC review will be based on, in part, the infrastructure readiness assessment, which is an assessment of industry's current ability to provide specified components for the FOAK NGNP, meet quality assurance requirements, transport components, have the necessary workforce in place, and have the necessary construction capabilities. AREVA and Westinghouse were contracted to perform independent assessments of industry's capabilities because of their experience with nuclear supply chains, which is a result of their experiences with the EPR and AP-1000 reactors. Both vendors produced infrastructure readiness assessment reports that identified key components and categorized these components into three groups based on their ability to be deployed in the FOAK plant. The NGNP project has several programs that are developing key components and capabilities. For these components, the NGNP project have provided input to properly assess the infrastructure readiness for these components.

  20. Microsoft Word - 51-9106032-001 NGNP Plant Design Requirements...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    NHS Nuclear Heat SourceNuclear Heat Supply NICS Nuclear Island Cooling Water System NIHS Nuclear Island HVAC System NOAK "N th "-of-a-Kind NQA Nuclear Quality Assurance NRC...

  1. NRC Licensing Strategy Development for the NGNP

    SciTech Connect (OSTI)

    Mark R. Holbrook; Trevor Cook

    2008-09-01T23:59:59.000Z

    The Next Generation Nuclear Plant (NGNP) project will provide the basis for commercialization of a new generation of advanced nuclear plants that utilize hightemperature gas-cooled reactor (HTGR) technology. The inherently safe HTGR design characteristics can be utilized to supply high-temperature process heat, co-generated electricity, and/or hydrogen for a number of industrial applications (e.g., petrochemical processes). Completion of the NGNP will result in a facility that demonstrates the safety and economics of the design, the commercial industrial potential of the technology, and the viability of the licensing strategy.

  2. NGNP Risk Management through Assessing Technology Readiness

    SciTech Connect (OSTI)

    John W. Collins

    2010-08-01T23:59:59.000Z

    Throughout the Next Generation Nuclear Plant (NGNP) project life cycle, technical risks are identified, analyzed, and mitigated and decisions are made regarding the design and selection of plant and sub-system configurations, components and their fabrication materials, and operating conditions. Risk resolution and decision making are key elements that help achieve project completion within budget and schedule constraints and desired plant availability. To achieve this objective, a formal decision-making and risk management process was developed for NGNP, based on proven systems engineering principles that have guided aerospace and military applications.

  3. Next Generation Nuclear Plant Materials Research and Development Program Plan

    SciTech Connect (OSTI)

    G. O. Hayner; E.L. Shaber

    2004-09-01T23:59:59.000Z

    The U.S Department of Energy (DOE) has selected the Very High Temperature Reactor (VHTR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production without greenhouse gas emissions. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed, thermal neutron spectrum reactor that will produce electricity and hydrogen in a state-of-the-art thermodynamically efficient manner. The NGNP will use very high burn-up, low-enriched uranium, TRISO-coated fuel and have a projected plant design service life of 60 years.

  4. NGNP Reactor Coolant Chemistry Control Study

    SciTech Connect (OSTI)

    Brian Castle

    2010-11-01T23:59:59.000Z

    The main focus of this paper is to identify the most desirable ranges of impurity levels in the primary coolant to optimize component life in the primary circuit of the Next Generation Nuclear Plant (NGNP), which will either be a prismatic block or pebble bed reactor.

  5. Assessment of NGNP Moisture Ingress Events

    SciTech Connect (OSTI)

    Bill Landman

    2011-04-01T23:59:59.000Z

    An assessment of modular HTGR moisture ingress events, making use of a phenomena identification and ranking process, was conducted by a panel of experts in the related areas for the U.S. next generation nuclear plant (NGNP) design. Consideration was given mainly to the prismatic core gas-cooled reactor configurations incorporating a steam generator within the primary circuit.

  6. NGNP Component Test Capability Design Code of Record

    SciTech Connect (OSTI)

    S.L. Austad; D.S. Ferguson; L.E. Guillen; C.W. McKnight; P.J. Petersen

    2009-09-01T23:59:59.000Z

    The Next Generation Nuclear Plant Project is conducting a trade study to select a preferred approach for establishing a capability whereby NGNP technology development testing—through large-scale, integrated tests—can be performed for critical HTGR structures, systems, and components (SSCs). The mission of this capability includes enabling the validation of interfaces, interactions, and performance for critical systems and components prior to installation in the NGNP prototype.

  7. NGNP Process Heat Applications: Hydrogen Production Accomplishments for FY2010

    SciTech Connect (OSTI)

    Charles V Park

    2011-01-01T23:59:59.000Z

    This report summarizes FY10 accomplishments of the Next Generation Nuclear Plant (NGNP) Engineering Process Heat Applications group in support of hydrogen production technology development. This organization is responsible for systems needed to transfer high temperature heat from a high temperature gas-cooled reactor (HTGR) reactor (being developed by the INL NGNP Project) to electric power generation and to potential industrial applications including the production of hydrogen.

  8. NGNP Graphite Selection and Acquisition Strategy

    SciTech Connect (OSTI)

    Burchell, T.; Bratton, R.; Windes, W.

    2007-09-30T23:59:59.000Z

    The nuclear graphite (H-451) previously used in the United States for High-Temperature Reactors (HTRs) is no longer available. New graphites have been developed and are considered suitable candidates for the Next-Generation Nuclear Plant (NGNP). A complete properties database for these new, available, candidate grades of graphite must be developed to support the design and licensing of NGNP core components. Data are required for the physical, mechanical (including radiation-induced creep), and oxidation properties of graphites. Moreover, the data must be statistically sound and take account of in-billet, between billets, and lot-to-lot variations of properties. These data are needed to support the ongoing development1 of the risk-derived American Society of Mechanical Engineers (ASME) graphite design code (a consensus code being prepared under the jurisdiction of the ASME by gas-cooled reactor and NGNP stakeholders including the vendors). The earlier Fort St. Vrain design of High-Temperature Reactor (HTRs) used deterministic performance models for H-451, while the NGNP will use new graphite grades and risk-derived (probabilistic) performance models and design codes, such as that being developed by the ASME. A radiation effects database must be developed for the currently available graphite materials, and this requires a substantial graphite irradiation program. The graphite Technology Development Plan (TDP)2 describes the data needed and the experiments planned to acquire these data in a timely fashion to support NGNP design, construction, and licensing. The strategy for the selection of appropriate grades of graphite for the NGNP is discussed here. The final selection of graphite grades depends upon the chosen reactor type and vendor because the reactor type (pebble bed or prismatic block) has a major influence on the graphite chosen by the designer. However, the time required to obtain the needed irradiation data for the selected NGNP graphite is sufficiently long that a preliminary selection was necessary in 2005. A further downselect was made in 2006, reducing the number of candidate graphites to two, with two reserve grades. Since then additional information has been obtained from potential NGNP vendors and graphite manufacturers. Therefore, the NGNP graphite selection strategy has been reassessed. New recommendations, and the rationale for these recommendations, are reported and discussed here.

  9. Next Generation Nuclear Plant Project Technology Development Roadmaps: The Technical Path Forward

    SciTech Connect (OSTI)

    John Collins

    2009-01-01T23:59:59.000Z

    This document presents the Next Generation Nuclear Plant (NGNP) Systems, Subsystems, and Components, establishes a baseline for the current technology readiness status, and provides a path forward to achieve increasing levels of technical maturity.

  10. Microsoft Word - NGNP _Work Plan Final_ - 13-Dec-06 _2_.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the risks to project completion. AREVA NGNP design team was established to bring in all core competencies necessary for NGNP design development from nuclear reactor and fuel...

  11. Next Generation Nuclear Plant Research and Development Program Plan

    SciTech Connect (OSTI)

    None

    2005-01-01T23:59:59.000Z

    The U.S Department of Energy (DOE) is conducting research and development (R&D) on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core could be either a prismatic graphite block type core or a pebble bed core. Use of a liquid salt coolant is also being evaluated. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The objectives of the NGNP Project are to: (1) Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission (2) Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, will perform R&D that will be critical to the success of the NGNP, primarily in the areas of: (1) High temperature gas reactor fuels behavior; (2) High temperature materials qualification; (3) Design methods development and validation; (4) Hydrogen production technologies; and (5) Energy conversion. The current R&D work is addressing fundamental issues that are relevant to a variety of possible NGNP designs. This document describes the NGNP R&D planned and currently underway in the first three topic areas listed above. The NGNP Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is presented in Section 2, the NGNP Materials R&D Program Plan is presented in Section 3, and the NGNP Design Methods Development and Validation R&D Program is presented in Section 4. The DOE-funded hydrogen production [DOE 2004] and energy conversion technologies programs are described elsewhere.

  12. Next Generation Nuclear Plant Research and Development Program Plan

    SciTech Connect (OSTI)

    P. E. MacDonald

    2005-01-01T23:59:59.000Z

    The U.S Department of Energy (DOE) is conducting research and development (R&D) on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core could be either a prismatic graphite block type core or a pebble bed core. Use of a liquid salt coolant is also being evaluated. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission Demonstrate safe and economical nuclearassisted production of hydrogen and electricity. The DOE laboratories, led by the INL, will perform R&D that will be critical to the success of the NGNP, primarily in the areas of: High temperature gas reactor fuels behavior High temperature materials qualification Design methods development and validation Hydrogen production technologies Energy conversion. The current R&D work is addressing fundamental issues that are relevant to a variety of possible NGNP designs. This document describes the NGNP R&D planned and currently underway in the first three topic areas listed above. The NGNP Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is presented in Section 2, the NGNP Materials R&D Program Plan is presented in Section 3, and the NGNP Design Methods Development and Validation R&D Program is presented in Section 4. The DOE-funded hydrogen production [DOE 2004] and energy conversion technologies programs are described elsewhere.

  13. Next Generation Nuclear Plant Project Evaluation of Siting a HTGR Co-generation Plant on an Operating Commercial Nuclear Power Plant Site

    SciTech Connect (OSTI)

    L.E. Demick

    2011-10-01T23:59:59.000Z

    This paper summarizes an evaluation by the Idaho National Laboratory (INL) Next Generation Nuclear Plant (NGNP) Project of siting a High Temperature Gas-cooled Reactor (HTGR) plant on an existing nuclear plant site that is located in an area of significant industrial activity. This is a co-generation application in which the HTGR Plant will be supplying steam and electricity to one or more of the nearby industrial plants.

  14. Updated NGNP Fuel Acquisition Strategy

    SciTech Connect (OSTI)

    David Petti; Tim Abram; Richard Hobbins; Jim Kendall

    2010-12-01T23:59:59.000Z

    A Next Generation Nuclear Plant (NGNP) fuel acquisition strategy was first established in 2007. In that report, a detailed technical assessment of potential fuel vendors for the first core of NGNP was conducted by an independent group of international experts based on input from the three major reactor vendor teams. Part of the assessment included an evaluation of the credibility of each option, along with a cost and schedule to implement each strategy compared with the schedule and throughput needs of the NGNP project. While credible options were identified based on the conditions in place at the time, many changes in the assumptions underlying the strategy and in externalities that have occurred in the interim requiring that the options be re-evaluated. This document presents an update to that strategy based on current capabilities for fuel fabrication as well as fuel performance and qualification testing worldwide. In light of the recent Pebble Bed Modular Reactor (PBMR) project closure, the Advanced Gas Reactor (AGR) fuel development and qualification program needs to support both pebble and prismatic options under the NGNP project. A number of assumptions were established that formed a context for the evaluation. Of these, the most important are: • Based on logistics associated with the on-going engineering design activities, vendor teams would start preliminary design in October 2012 and complete in May 2014. A decision on reactor type will be made following preliminary design, with the decision process assumed to be completed in January 2015. Thus, no fuel decision (pebble or prismatic) will be made in the near term. • Activities necessary for both pebble and prismatic fuel qualification will be conducted in parallel until a fuel form selection is made. As such, process development, fuel fabrication, irradiation, and testing for pebble and prismatic options should not negatively influence each other during the period prior to a decision on reactor type. • Additional funding will be made available beginning in fiscal year (FY) 2012 to support pebble bed fuel fabrication process development and fuel testing while maintaining the prismatic fuel schedule. Options for fuel fabrication for prismatic and pebble bed were evaluated based on the credibility of each option, along with a cost and schedule to implement each strategy. The sole prismatic option is Babcock and Wilcox (B&W) producing uranium oxycarbide (UCO) tristructural-isotropic (TRISO) fuel particles in compacts. This option finishes in the middle of 2022 . Options for the pebble bed are Nuclear Fuel Industries (NFI) in Japan producing uranium dioxide (UO2) TRISO fuel particles, and/or B&W producing UCO or UO2 TRISO fuel particles. All pebble options finish in mid to late 2022.

  15. NGNP Risk Management Database: A Model for Managing Risk

    SciTech Connect (OSTI)

    John Collins

    2009-09-01T23:59:59.000Z

    To facilitate the implementation of the Risk Management Plan, the Next Generation Nuclear Plant (NGNP) Project has developed and employed an analytical software tool called the NGNP Risk Management System (RMS). A relational database developed in Microsoft® Access, the RMS provides conventional database utility including data maintenance, archiving, configuration control, and query ability. Additionally, the tool’s design provides a number of unique capabilities specifically designed to facilitate the development and execution of activities outlined in the Risk Management Plan. Specifically, the RMS provides the capability to establish the risk baseline, document and analyze the risk reduction plan, track the current risk reduction status, organize risks by reference configuration system, subsystem, and component (SSC) and Area, and increase the level of NGNP decision making.

  16. Next Generation Nuclear Plant Materials Research and Development Program Plan

    SciTech Connect (OSTI)

    G.O. Hayner; R.L. Bratton; R.N. Wright

    2005-09-01T23:59:59.000Z

    The U.S Department of Energy (DOE) has selected the Very High Temperature Reactor (VHTR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production without greenhouse gas emissions. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed, thermal neutron spectrum reactor that will produce electricity and hydrogen in a state-of-the-art thermodynamically efficient manner. The NGNP will use very high burn-up, low-enriched uranium, TRISO-coated fuel and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Project is envisioned to demonstrate the following: (1) A full-scale prototype VHTR by about 2021; (2) High-temperature Brayton Cycle electric power production at full scale with a focus on economic performance; (3) Nuclear-assisted production of hydrogen (with about 10% of the heat) with a focus on economic performance; and (4) By test, the exceptional safety capabilities of the advanced gas-cooled reactors. Further, the NGNP program will: (1) Obtain a Nuclear Regulatory Commission (NRC) License to construct and operate the NGNP, this process will provide a basis for future performance based, risk-informed licensing; and (2) Support the development, testing, and prototyping of hydrogen infrastructures. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. The NGNP Materials R&D Program includes the following elements: (1) Developing a specific approach, program plan and other project management tools for managing the R&D program elements; (2) Developing a specific work package for the R&D activities to be performed during each government fiscal year; (3) Reporting the status and progress of the work based on committed deliverables and milestones; (4) Developing collaboration in areas of materials R&D of benefit to the NGNP with countries that are a part of the Generation IV International Forum; and (5) Ensuring that the R&D work performed in support of the materials program is in conformance with established Quality Assurance and procurement requirements. The objective of the NGNP Materials R&D Program is to provide the essential materials R&D needed to support the design and licensing of the reactor and balance of plant, excluding the hydrogen plant. The materials R&D program is being initiated prior to the design effort to ensure that materials R&D activities are initiated early enough to support the design process and support the Project Integrator. The thermal, environmental, and service life conditions of the NGNP will make selection and qualification of some high-temperature materials a significant challenge; thus, new materials and approaches may be required.

  17. The Next Generation Nuclear Plant

    SciTech Connect (OSTI)

    Dr. David A. Petti

    2009-01-01T23:59:59.000Z

    The Next Generation Nuclear Plant (NGNP) will be a demonstration of the technical, licensing, operational, and commercial viability of High Temperature Gas-Cooled Reactor (HTGR) technology for the production of process heat, electricity, and hydrogen. This nuclear- based technology can provide high-temperature process heat (up to 950°C) that can be used as a substitute for the burning of fossil fuels for a wide range of commercial applications (see Figure 1). The substitution of the HTGR for burning fossil fuels conserves these hydrocarbon resources for other uses, reduces uncertainty in the cost and supply of natural gas and oil, and eliminates the emissions of greenhouse gases attendant with the burning of these fuels. The HTGR is a passively safe nuclear reactor concept with an easily understood safety basis that permits substantially reduced emergency planning requirements and improved siting flexibility compared to other nuclear technologies.

  18. NGNP SITE 2 HAZARDS ASSESSMENT

    SciTech Connect (OSTI)

    Wayne Moe

    2011-10-01T23:59:59.000Z

    The Next Generation Nuclear Plant (NGNP) Project initiated at Idaho National Laboratory (INL) by the U.S. Department of Energy pursuant to the 2005 Energy Policy Act, is based on research and development activities supported by the Generation IV Nuclear Energy Systems Initiative. The principal objective of the NGNP Project is to support commercialization of the high temperature gas-cooled reactor (HTGR) technology. The HTGR is a helium-cooled and graphite-moderated reactor that can operate at temperatures much higher than those of conventional light water reactor (LWR) technologies. Accordingly, it can be applied in many industrial applications as a substitute for burning fossil fuels, such as natural gas, to generate process heat in addition to producing electricity, which is the principal application of current LWRs. Nuclear energy in the form of LWRs has been used in the U.S. and internationally principally for the generation of electricity. However, because the HTGR operates at higher temperatures than LWRs, it can be used to displace the use of fossil fuels in many industrial applications. It also provides a carbon emission-free energy supply. For example, the energy needs for the recovery and refining of petroleum, for the petrochemical industry and for production of transportation fuels and feedstocks using coal conversion processes require process heat provided at temperatures approaching 800 C. This temperature range is readily achieved by the HTGR technology. This report summarizes a site assessment authorized by INL under the NGNP Project to determine hazards and potential challenges that site owners and HTGR designers need to be aware of when developing the HTGR design for co-location at industrial facilities, and to evaluate the site for suitability considering certain site characteristics. The objectives of the NGNP site hazard assessments are to do an initial screening of representative sites in order to identify potential challenges and restraints to be addressed in design and licensing processes; assure the HTGR technology can be deployed at variety of sites for a range of applications; evaluate potential sites for potential hazards and describe some of the actions necessary to mitigate impacts of hazards; and, provide key insights that can inform the plant design process. The report presents a summary of the process methodology and the results of an assessment of hazards typical of a class of candidate sites for the potential deployment of HTGR reactor technology. The assessment considered health and safety, and other important siting characteristics to determine the potential impact of identified hazards and potential challenges presented by the location for this technology. A four reactor module nuclear plant (2000 to 2400 MW thermal), that co-generates steam, electricity for general use in the plant, and hot gas for use in a nearby chemical processing facility, to provide the requisite performance and reliability was assumed for the assessment.

  19. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    SciTech Connect (OSTI)

    J. K. Wright; R. N. Wright

    2010-07-01T23:59:59.000Z

    The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750°C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

  20. Next Generation Nuclear Plant GAP Analysis Report

    SciTech Connect (OSTI)

    Ball, Sydney J [ORNL; Burchell, Timothy D [ORNL; Corwin, William R [ORNL; Fisher, Stephen Eugene [ORNL; Forsberg, Charles W. [Massachusetts Institute of Technology (MIT); Morris, Robert Noel [ORNL; Moses, David Lewis [ORNL

    2008-12-01T23:59:59.000Z

    As a follow-up to the phenomena identification and ranking table (PIRT) studies conducted recently by NRC on next generation nuclear plant (NGNP) safety, a study was conducted to identify the significant 'gaps' between what is needed and what is already available to adequately assess NGNP safety characteristics. The PIRT studies focused on identifying important phenomena affecting NGNP plant behavior, while the gap study gives more attention to off-normal behavior, uncertainties, and event probabilities under both normal operation and postulated accident conditions. Hence, this process also involved incorporating more detailed evaluations of accident sequences and risk assessments. This study considers thermal-fluid and neutronic behavior under both normal and postulated accident conditions, fission product transport (FPT), high-temperature metals, and graphite behavior and their effects on safety. In addition, safety issues related to coupling process heat (hydrogen production) systems to the reactor are addressed, given the limited design information currently available. Recommendations for further study, including analytical methods development and experimental needs, are presented as appropriate in each of these areas.

  1. NGNP PHASE I REVIEW

    Broader source: Energy.gov (indexed) [DOE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Careerlumens_placard-green.eps MoreWSRC-STI-2007-00250This NuclearDepartmentNGNP PHASE I REVIEW NEAC

  2. Design Features and Technology Uncertainties for the Next Generation Nuclear Plant

    SciTech Connect (OSTI)

    John M. Ryskamp; Phil Hildebrandt; Osamu Baba; Ron Ballinger; Robert Brodsky; Hans-Wolfgang Chi; Dennis Crutchfield; Herb Estrada; Jeane-Claude Garnier; Gerald Gordon; Richard Hobbins; Dan Keuter; Marilyn Kray; Philippe Martin; Steve Melancon; Christian Simon; Henry Stone; Robert Varrin; Werner von Lensa

    2004-06-01T23:59:59.000Z

    This report presents the conclusions, observations, and recommendations of the Independent Technology Review Group (ITRG) regarding design features and important technology uncertainties associated with very-high-temperature nuclear system concepts for the Next Generation Nuclear Plant (NGNP). The ITRG performed its reviews during the period November 2003 through April 2004.

  3. Microsoft PowerPoint - Project Briefing for Nuclear Energy Advisory...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Next Generation Nuclear Plant Next Generation Nuclear Plant . Project Briefing for . Nuclear Energy Advisory Committee uc ea e gy d so y Co ttee Greg Gibbs Director NGNP Project...

  4. NGNP Process Heat Utilization: Liquid Metal Phase Change Heat Exchanger

    SciTech Connect (OSTI)

    Piyush Sabharwall; Mike Patterson; Vivek Utgikar; Fred Gunnerson

    2008-09-01T23:59:59.000Z

    One key long-standing issue that must be overcome to fully realize the successful growth of nuclear power is to determine other benefits of nuclear energy apart from meeting the electricity demands. The Next Generation Nuclear Plant (NGNP) will most likely be producing electricity and heat for the production of hydrogen and/or oil retrieval from oil sands and oil shale to help in our national pursuit of energy independence. For nuclear process heat to be utilized, intermediate heat exchange is required to transfer heat from the NGNP to the hydrogen plant or oil recovery field in the most efficient way possible. Development of nuclear reactor - process heat technology has intensified the interest in liquid metals as heat transfer media because of their ideal transport properties. Liquid metal heat exchangers are not new in practical applications. An important rational for considering liquid metals is the potential convective heat transfer is among the highest known. Thus explains the interest in liquid metals as coolant for intermediate heat exchange from NGNP. For process heat it is desired that, intermediate heat exchangers (IHX) transfer heat from the NGNP in the most efficient way possible. The production of electric power at higher efficiency via the Brayton Cycle, and hydrogen production, requires both heat at higher temperatures and high effectiveness compact heat exchangers to transfer heat to either the power or process cycle. Compact heat exchangers maximize the heat transfer surface area per volume of heat exchanger; this has the benefit of reducing heat exchanger size and heat losses. High temperature IHX design requirements are governed in part by the allowable temperature drop between the outlet and inlet of the NGNP. In order to improve the characteristics of heat transfer, liquid metal phase change heat exchangers may be more effective and efficient. This paper explores the overall heat transfer characteristics and pressure drop of the phase change heat exchanger with Na as the heat exchanger coolant. In order to design a very efficient and effective heat exchanger one must optimize the design such that we have a high heat transfer and a lower pressure drop, but there is always a trade-off between them. Based on NGNP operational parameters, a heat exchanger analysis with the sodium phase change will be presented to show that the heat exchanger has the potential for highly effective heat transfer, within a small volume at reasonable cost.

  5. NGNP – Creating Validated TRL and TDRMs for Critical Systems, Subsystems, and Components

    SciTech Connect (OSTI)

    John W. Collins; John M. Beck; Emmanuel O. Opare; Layne F. Pincock

    2008-09-01T23:59:59.000Z

    This report introduces two draft Next Generation Nuclear Plant (NGNP) Technology Development Roadmaps (TDRMs) and documents the methods used to create them. As such, this report depicts the development of the hardware needed to successfully operate the NGNP and identifies this hardware by the area of the plant it supports and by system, subsystem, and component (SSC). Several options exist for which technologies are selected to fulfill the functions of the NGNP. These options are represented by differing SSCs and are grouped into reference designs. Each SSC associated with each reference design is evaluated, rated, and assigned a technology readiness level (TRL). A rollup of the TRLs allows for comparison of the various reference designs. A TDRM then documents the tasks needed to obtain information in key discriminating criteria to support technology down selection and the tasks and test required to sufficiently mature the technology and reduce the likelihood of technological failure upon installation. This report presents the path forward, methods, and tools used to understand the requirements, manage the uncertainty, and mitigate the risk for the NGNP project. The key tool, TDRMs, is the means to facilitate NGNP risk-informed decision making, technology down selection, and technology qualification and maturation while serving to coordinate engineering, research and development, and licensing efforts.

  6. NGNP Risk Management Database: A Model for Managing Risk

    SciTech Connect (OSTI)

    John Collins; John M. Beck

    2011-11-01T23:59:59.000Z

    The Next Generation Nuclear Plant (NGNP) Risk Management System (RMS) is a database used to maintain the project risk register. The RMS also maps risk reduction activities to specific identified risks. Further functionality of the RMS includes mapping reactor suppliers Design Data Needs (DDNs) to risk reduction tasks and mapping Phenomena Identification Ranking Table (PIRTs) to associated risks. This document outlines the basic instructions on how to use the RMS. This document constitutes Revision 1 of the NGNP Risk Management Database: A Model for Managing Risk. It incorporates the latest enhancements to the RMS. The enhancements include six new custom views of risk data - Impact/Consequence, Tasks by Project Phase, Tasks by Status, Tasks by Project Phase/Status, Tasks by Impact/WBS, and Tasks by Phase/Impact/WBS.

  7. Summary of Differences in Approach to Executing the NGNP Project

    SciTech Connect (OSTI)

    P.C. Hildebrandt

    2010-08-01T23:59:59.000Z

    The following discuss the summary differences between the NGNP Industry Alliance’s approach to execution of the NGNP Project as described in its Project Execution Strategy, and that included in the DOE’s Report to Congress on the Next Generation Nuclear Plant dated April 2010. These summary differences are not a comprehensive description of the results of a detailed gap analysis, but rather are those differences that warrant discussion in meetings between DOE executives and the private sector represented by the Alliance. In practical fact, the full range and content of the differences will not be understood until such time as detailed discussions are engaged between the DOE and the Alliance to develop a mutually agreed-to project execution plan.

  8. NUCLEAR PLANT OPERATIONS AND

    E-Print Network [OSTI]

    Pázsit, Imre

    NUCLEAR PLANT OPERATIONS AND CONTROL KEYWORDS: neutron flux, cur- rent noise, vibration diagnostics: Swedish Nuclear Powe

  9. Summary of Planned Implementation for the HTGR Lessons Learned Applicable to the NGNP

    SciTech Connect (OSTI)

    Ian Mckirdy

    2011-09-01T23:59:59.000Z

    This document presents a reconciliation of the lessons learned during a 2010 comprehensive evaluation of pertinent lessons learned from past and present high temperature gas-cooled reactors that apply to the Next Generation Nuclear Plant Project along with current and planned activities. The data used are from the latest Idaho National Laboratory research and development plans, the conceptual design report from General Atomics, and the pebble bed reactor technology readiness study from AREVA. Only those lessons related to the structures, systems, and components of the Next Generation Nuclear Plant (NGNP), as documented in the recently updated lessons learned report are addressed. These reconciliations are ordered according to plant area, followed by the affected system, subsystem, or component; lesson learned; and finally an NGNP implementation statement. This report (1) provides cross references to the original lessons learned document, (2) describes the lesson learned, (3) provides the current NGNP implementation status with design data needs associated with the lesson learned, (4) identifies the research and development being performed related to the lesson learned, and (5) summarizes with a status of how the lesson learned has been addressed by the NGNP Project.

  10. RELAP5-3D Transient Modelling for NGNP Integrated Plant

    SciTech Connect (OSTI)

    P. Sabharwall; N.A. Anderson

    2014-06-01T23:59:59.000Z

    The High-Temperature Gas-cooled Reactor (HTGR) is designed with outlet temperatures ranging between 750°C and 800°C. These high outlet temperatures enhance the power production efficiency and facilitate a variety of industrial applications. The objective of this study is to understand the response of the primary system to potential transients in the secondary system. For this analysis, the transient condition originates in the Intermediate Heat Exchanger (IHX) or Steam Generator (SG) of the HTGR-integrated plant. The transients analysed are: a loss of pressure; loss of feedwater flow; inadvertent closure of main steam valve; decrease in returning gas temperature and heat load step change. The results show a large dependence on the negative reactivity added to the fuel as a function of increased temperature. The returning gas temperature decrease transient resulted in the highest fuel temperature (1361°C). Fuel temperature was shown to be less than the 1600°C fuel limit for each case analysed.

  11. TRITIUM BARRIER MATERIALS AND SEPARATION SYSTEMS FOR THE NGNP

    SciTech Connect (OSTI)

    Sherman, S; Thad Adams, T

    2008-07-17T23:59:59.000Z

    Contamination of downstream hydrogen production plants or other users of high-temperature heat is a concern of the Next Generation Nuclear Plant (NGNP) Project. Due to the high operating temperatures of the NGNP (850-900 C outlet temperature), tritium produced in the nuclear reactor can permeate through heat exchangers to reach the hydrogen production plant, where it can become incorporated into process chemicals or the hydrogen product. The concentration limit for tritium in the hydrogen product has not been established, but it is expected that any future limit on tritium concentration will be no higher than the air and water effluent limits established by the NRC and the EPA. A literature survey of tritium permeation barriers, capture systems, and mitigation measures is presented and technologies are identified that may reduce the movement of tritium to the downstream plant. Among tritium permeation barriers, oxide layers produced in-situ may provide the most suitable barriers, though it may be possible to use aluminized surfaces also. For tritium capture systems, the use of getters is recommended, and high-temperature hydride forming materials such as Ti, Zr, and Y are suggested. Tritium may also be converted to HTO in order to capture it on molecular sieves or getter materials. Counter-flow of hydrogen may reduce the flux of tritium through heat exchangers. Recommendations for research and development work are provided.

  12. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    SciTech Connect (OSTI)

    J. K. Wright; R. N. Wright

    2008-04-01T23:59:59.000Z

    The U.S. Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic, or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development Program is responsible for performing research and development on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2¼Cr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have concluded, however, that with adequate engineered cooling of the vessel, the A508/533 steels are also acceptable.

  13. Microsoft Word - NGNP_Licensing Study Report_4-30-2008-FINAL...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Company LLC 8 of 55 Abbreviation or Acronym Definition MHTGR Modular High Temperature Gas-Cooled Reactor NEPA National Environmental Policy Act NGNP Next Generation Nuclear...

  14. Reactor Physics Parametric and Depletion Studies in Support of TRISO Particle Fuel Specification for the Next Generation Nuclear Plant

    SciTech Connect (OSTI)

    James W. Sterbentz; Bren Phillips; Robert L. Sant; Gray S. Chang; Paul D. Bayless

    2003-09-01T23:59:59.000Z

    Reactor physics calculations were initiated to answer several major questions related to the proposed TRISO-coated particle fuel that is to be used in the prismatic Very High Temperature Reactor (VHTR) or the Next Generation Nuclear Plant (NGNP). These preliminary design evaluation calculations help ensure that the upcoming fuel irradiation tests will test appropriate size and type of fuel particles for a future NGNP reactor design. Conclusions from these calculations are expected to confirm and suggest possible modifications to the current particle fuel parameters specified in the evolving Fuel Specification. Calculated results dispel the need for a binary fuel particle system, which is proposed in the General Atomics GT-MHR concept. The GT-MHR binary system is composed of both a fissile and fertile particle with 350- and 500- micron kernel diameters, respectively. For the NGNP reactor, a single fissile particle system (single UCO kernel size) can meet the reactivity and power cycle length requirements demanded of the NGNP. At the same time, it will provide substantial programmatic cost savings by eliminating the need for dual particle fabrication process lines and dual fuel particle irradiation tests required of a binary system. Use of a larger 425-micron kernel diameter single fissile particle (proposed here), as opposed to the 350-micron GT-MHR fissile particle size, helps alleviate current compact particle packing fractions fabrication limitations (<35%), improves fuel block loading for higher n-batch reload options, and tracks the historical correlation between particle size and enrichment (10 and 14 wt% U-235 particle enrichments are proposed for the NGNP). Overall, the use of the slightly larger kernel significantly broadens the NGNP reactor core design envelope and provides increased design margin to accommodate the (as yet) unknown final NGNP reactor design. Maximum power-peaking factors are calculated for both the initial and equilibrium NGNP cores. Radial power-peaking can be fully controlled with particle packing fraction zoning (no enrichment zoning required) and discrete burnable poison rods. Optimally loaded NGNP cores can expect radial powerpeaking factors as low as 1.14 at beginning of cycle (BOC), increasing slowly to a value of 1.25 by end of cycle (EOC), an axial power-peaking value of 1.30 (BOC), and for individual fuel particles in the maximum compact <1.05 (BOC) and an approximate value of 1.20 (EOC) due to Pu-239 buildup in particles on the compact periphery. The NGNP peak particle powers, using a conservative total power-peaking factor (~2.1 factor), are expected to be <150 mW/particle (well below the 250 mW/particle limit, even with the larger 425-micron kernel size).

  15. Next Generation Nuclear Plant Intermediate Heat Exchanger Materials Research and Development Plan (PLN-2804)

    SciTech Connect (OSTI)

    J. K. Wright

    2008-04-01T23:59:59.000Z

    DOE has selected the High Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Today’s high-temperature alloys and associated ASME Codes for reactor applications are approved up to 760°C. However, some primary system components, such as the Intermediate Heat Exchanger (IHX) for the NGNP will require use of materials that can withstand higher temperatures. The thermal, environmental, and service life conditions of the NGNP will make selection and qualification of some high-temperature materials a significant challenge. Examples include materials for the core barrel and core internals, such as the control rod sleeves. The requirements of the materials for the IHX are among the most demanding. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while at the same time setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. A number of solid solution strengthened nickel based alloys have been considered for application in heat exchangers and core internals for the NGNP. The primary candidates are Inconel 617, Haynes 230, Incoloy 800H and Hastelloy XR. Based on the technical maturity, availability in required product forms, experience base, and high temperature mechanical properties all of the vendor pre-conceptual design studies have specified Alloy 617 as the material of choice for heat exchangers. Also a draft code case for Alloy 617 was developed previously. Although action was suspended before the code case was accepted by ASME, this draft code case provides a significant head start for achieving codification of the material. Similarly, Alloy 800H is the material of choice for control rod sleeves. In addition to the above listed considerations, Alloy 800H is already listed in the nuclear section of the ASME Code; although the maximum use temperature and time need to be increased.

  16. NGNP Project 2011 Status and Path Forward

    SciTech Connect (OSTI)

    L.E. Demick

    2011-12-01T23:59:59.000Z

    High Temperature Gas Reactor (HTGR) technology can play an important role in the United States’ energy future by extending the use of nuclear energy for non-electricity energy production missions as well as continuing to provide a considerable base load electric power generation capability. Extending nuclear energy into the industrial and transportation sectors through the co-production of process heat and electricity provides safe and reliable energy for these sectors in an environmentally responsible manner. The safety case for the modular HTGR provides a substantial improvement in nuclear plant safety for the protection of the public and the environment, and supports collocation of the HTGR with major industrial facilities. The NGNP Project at the Idaho National Laboratory has been working toward an objective of commercializing the HTGR technology under DOE direction since 2006. The Project is undergoing a quantum shift in direction and scope as a result of recent DOE decisions. This paper summarizes where the Project has been, where it is at the time of this writing and what is needed in future activities to commercialize HTGR technology.

  17. Balance of Plant Requirements for a Nuclear Hydrogen Plant

    SciTech Connect (OSTI)

    Bradley Ward

    2006-04-01T23:59:59.000Z

    This document describes the requirements for the components and systems that support the hydrogen production portion of a 600 megawatt thermal (MWt) Next Generation Nuclear Plant (NGNP). These systems, defined as the "balance-of-plant" (BOP), are essential to operate an effective hydrogen production plant. Examples of BOP items are: heat recovery and heat rejection equipment, process material transport systems (pumps, valves, piping, etc.), control systems, safety systems, waste collection and disposal systems, maintenance and repair equipment, heating, ventilation, and air conditioning (HVAC), electrical supply and distribution, and others. The requirements in this document are applicable to the two hydrogen production processes currently under consideration in the DOE Nuclear Hydrogen Initiative. These processes are the sulfur iodide (S-I) process and the high temperature electrolysis (HTE) process. At present, the other two hydrogen production process - the hybrid sulfur-iodide electrolytic process (SE) and the calcium-bromide process (Ca-Br) -are under flow sheet development and not included in this report. While some features of the balance-of-plant requirements are common to all hydrogen production processes, some details will apply only to the specific needs of individual processes.

  18. NUCLEAR PLANT AND CONTROL

    E-Print Network [OSTI]

    NUCLEAR PLANT OPERATIONS AND CONTROL KEYWORDS: software require- ments, safety analysis, formal for the digital protection systems of a nuclear power plant. When spec- ifying requirements for software and CRSA processes are described using shutdown system 2 of the Wolsong nuclear power plants as the digital

  19. NGNP Project Regulatory Gap Analysis for Modular HTGRs

    SciTech Connect (OSTI)

    Wayne Moe

    2011-09-01T23:59:59.000Z

    The Next Generation Nuclear Plant (NGNP) Project Regulatory Gap Analysis (RGA) for High Temperature Gas-Cooled Reactors (HTGR) was conducted to evaluate existing regulatory requirements and guidance against the design characteristics specific to a generic modular HTGR. This final report presents results and identifies regulatory gaps concerning current Nuclear Regulatory Commission (NRC) licensing requirements that apply to the modular HTGR design concept. This report contains appendices that highlight important HTGR licensing issues that were found during the RGA study. The information contained in this report will be used to further efforts in reconciling HTGR-related gaps in the NRC licensing structure, which has to date largely focused on light water reactor technology.

  20. NUCLEAR PLANT OPERATIONS AND

    E-Print Network [OSTI]

    Demazière, Christophe

    NUCLEAR PLANT OPERATIONS AND CONTROL KEYWORDS: moderator temper ature coefficient, reactivity co reactor Unit 4 of the Ringhals Nuclear Power Plant (Sweden) during fuel cycle 16 is analyzed absorption cross-section behavior. Consequently, if NUCLEAR TECHNOLOGY VOL. 140 NOV. 2002 147 #12;Demazière

  1. NUCLEAR PLANT OPERATIONS AND

    E-Print Network [OSTI]

    Pázsit, Imre

    NUCLEAR PLANT OPERATIONS AND CONTROL KEYWORDS: moderator temper- ature coefficient, reactivity co reactor Unit 4 of the Ringhals Nuclear Power Plant (Sweden) during fuel cycle 16 is analyzed. Consequently, if*E-mail: demaz@nephy.chalmers.se NUCLEAR TECHNOLOGY VOL. 140 NOV. 2002 147 #12;high-burnup fuel

  2. Next Generation Nuclear Plant Project 2009 Status Report

    SciTech Connect (OSTI)

    Larry Demick; Jim Kinsey; Keith Perry; Dave Petti

    2010-05-01T23:59:59.000Z

    The mission of the NGNP Project is to broaden the environmental and economic benefits of nuclear energy technology to the United States and other economies by demonstrating its applicability to market sectors not served by light water reactors (LWRs). Those markets typically use fossil fuels to fulfill their energy needs, and high temperature gas-cooled reactors (HTGRs) like the NGNP can reduce this dependence and the resulting carbon footprint.

  3. Hydrogen Production from the Next Generation Nuclear Plant

    SciTech Connect (OSTI)

    M. Patterson; C. Park

    2008-03-01T23:59:59.000Z

    The Next Generation Nuclear Plant (NGNP) is a high temperature gas-cooled reactor that will be capable of producing hydrogen, electricity and/or high temperature process heat for industrial use. The project has initiated the conceptual design phase and when completed will demonstrate the viability of hydrogen generation using nuclear produced process heat. This paper explains how industry and the U.S. Government are cooperating to advance nuclear hydrogen technology. It also describes the issues being explored and the results of recent R&D including materials development and testing, thermal-fluids research, and systems analysis. The paper also describes the hydrogen production technologies being considered (including various thermochemical processes and high-temperature electrolysis).

  4. Research and Development Technology Development Roadmaps for the Next Generation Nuclear Plant Project

    SciTech Connect (OSTI)

    Ian McKirdy

    2011-07-01T23:59:59.000Z

    The U.S. Department of Energy (DOE) has selected the high temperature gas-cooled reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for process heat, hydrogen and electricity production. The reactor will be graphite moderated with helium as the primary coolant and may be either prismatic or pebble-bed. Although, final design features have not yet been determined. Research and Development (R&D) activities are proceeding on those known plant systems to mature the technology, codify the materials for specific applications, and demonstrate the component and system viability in NGNP relevant and integrated environments. Collectively these R&D activities serve to reduce the project risk and enhance the probability of on-budget, on-schedule completion and NRC licensing. As the design progresses, in more detail, toward final design and approval for construction, selected components, which have not been used in a similar application, in a relevant environment nor integrated with other components and systems, must be tested to demonstrate viability at reduced scales and simulations prior to full scale operation. This report and its R&D TDRMs present the path forward and its significance in assuring technical readiness to perform the desired function by: Choreographing the integration between design and R&D activities; and proving selected design components in relevant applications.

  5. Progress Report for Diffusion Welding of the NGNP Process Application Heat Exchangers

    SciTech Connect (OSTI)

    R.E. Mizia; D.E. Clark; M.V. Glazoff; T.E. Lister; T.L. Trowbridge

    2011-12-01T23:59:59.000Z

    The U.S. Department of Energy selected the high temperature gas-cooled reactor as the basis for the Next Generation Nuclear Plant (NGNP). The NGNP will demonstrate the use of nuclear power for electricity, hydrogen production, and process heat applications. The NGNP Project is currently investigating the use of metallic, diffusion welded, compact heat exchangers to transfer heat from the primary (reactor side) heat transport system to the secondary heat transport system. An intermediate heat exchanger will transfer this heat to downstream applications such as hydrogen production, process heat, and electricity generation. The channeled plates that make up the heat transfer surfaces of the intermediate heat exchanger will have to be assembled into an array by diffusion welding. This report describes the preliminary results of a scoping study that evaluated the diffusion welding process parameters and the resultant mechanical properties of diffusion welded joints using Alloy 800H. The long-term goal of the program is to progress towards demonstration of small heat exchanger unit cells fabricated with diffusion welds. Demonstration through mechanical testing of the unit cells will support American Society of Mechanical Engineers rules and standards development, reduce technical risk, and provide proof of concept for heat exchanger fabrication methods needed to deploy heat exchangers in several potential NGNP configurations.1 Researchers also evaluated the usefulness of modern thermodynamic and diffusion computational tools (Thermo-Calc and Dictra) in optimizing the parameters for diffusion welding of Alloy 800H. The modeling efforts suggested a temperature of 1150 C for 1 hour with an applied pressure of 5 MPa using 15 {micro}m nickel foil as joint filler to reduce chromium oxidation on the welded surfaces. Good agreement between modeled and experimentally determined concentration gradients was achieved

  6. Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 6: Process Heat and Hydrogen Co-Generation PIRTs

    SciTech Connect (OSTI)

    Forsberg, Charles W [ORNL; Gorensek, M. B. [Savannah River National Laboratory (SRNL); Herring, S. [Idaho National Laboratory (INL); Pickard, P. [Sandia National Laboratories (SNL)

    2008-03-01T23:59:59.000Z

    A Phenomena Identification and Ranking Table (PIRT) exercise was conducted to identify potential safety-0-related physical phenomena for the Next Generation Nuclear Plant (NGNP) when coupled to a hydrogen production or similar chemical plant. The NGNP is a very high-temperature reactor (VHTR) with the design goal to produce high-temperature heat and electricity for nearby chemical plants. Because high-temperature heat can only be transported limited distances, the two plants will be close to each other. One of the primary applications for the VHTR would be to supply heat and electricity for the production of hydrogen. There was no assessment of chemical plant safety challenges. The primary application of this PIRT is to support the safety analysis of the NGNP coupled one or more small hydrogen production pilot plants. However, the chemical plant processes to be coupled to the NGNP have not yet been chosen; thus, a broad PIRT assessment was conducted to scope alternative potential applications and test facilities associated with the NGNP. The hazards associated with various chemicals and methods to minimize risks from those hazards are well understood within the chemical industry. Much but not all of the information required to assure safe conditions (separation distance, relative elevation, berms) is known for a reactor coupled to a chemical plant. There is also some experience with nuclear plants in several countries that have produced steam for industrial applications. The specific characteristics of the chemical plant, site layout, and the maximum stored inventories of chemicals can provide the starting point for the safety assessments. While the panel identified events and phenomena of safety significance, there is one added caveat. Multiple high-temperature reactors provide safety-related experience and understanding of reactor safety. In contrast, there have been only limited safety studies of coupled chemical and nuclear plants. The work herein provides a starting point for those studies; but, the general level of understanding of safety in coupling nuclear and chemical plants is less than in other areas of high-temperature reactor safety.

  7. Next Generation Nuclear Plant Phenomena

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    and passive heat dissipation to withstand design basis events with minimal fuel damage and source term generation. As such, the NGNP places a burden on the designer to...

  8. Virginia Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

  9. Ohio Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    Ohio nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

  10. Arkansas Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

  11. Michigan Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

  12. Alabama Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

  13. Texas Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

  14. Tennessee Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    Tennessee nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

  15. Georgia Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

  16. Nebraska Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    Nebraska nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

  17. Arizona Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

  18. Maryland Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

  19. Illinois Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    Illinois nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

  20. Florida Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    Florida nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

  1. Wisconsin Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    Wisconsin nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

  2. Minnesota Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    Minnesota nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

  3. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

    SciTech Connect (OSTI)

    Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

    2007-03-21T23:59:59.000Z

    In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for both the PBMR and prismatic design. The main focus of this report is the RPV for both design concepts with emphasis on material selection.

  4. Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 5: Graphite PIRTs

    SciTech Connect (OSTI)

    Burchell, Timothy D [ORNL; Bratton, Rob [Idaho National Laboratory (INL); Marsden, Barry [University of Manchester, UK; Srinivasan, Makuteswara [U.S. Nuclear Regulatory Commission; Penfield, Scott [Technology Insights; Mitchell, Mark [PBMR (Pty) Ltd.; Windes, Will [Idaho National Laboratory (INL)

    2008-03-01T23:59:59.000Z

    Here we report the outcome of the application of the Nuclear Regulatory Commission (NRC) Phenomena Identification and Ranking Table (PIRT) process to the issue of nuclear-grade graphite for the moderator and structural components of a next generation nuclear plant (NGNP), considering both routine (normal operation) and postulated accident conditions for the NGNP. The NGNP is assumed to be a modular high-temperature gas-cooled reactor (HTGR), either a gas-turbine modular helium reactor (GTMHR) version [a prismatic-core modular reactor (PMR)] or a pebble-bed modular reactor (PBMR) version [a pebble bed reactor (PBR)] design, with either a direct- or indirect-cycle gas turbine (Brayton cycle) system for electric power production, and an indirect-cycle component for hydrogen production. NGNP design options with a high-pressure steam generator (Rankine cycle) in the primary loop are not considered in this PIRT. This graphite PIRT was conducted in parallel with four other NRC PIRT activities, taking advantage of the relationships and overlaps in subject matter. The graphite PIRT panel identified numerous phenomena, five of which were ranked high importance-low knowledge. A further nine were ranked with high importance and medium knowledge rank. Two phenomena were ranked with medium importance and low knowledge, and a further 14 were ranked medium importance and medium knowledge rank. The last 12 phenomena were ranked with low importance and high knowledge rank (or similar combinations suggesting they have low priority). The ranking/scoring rationale for the reported graphite phenomena is discussed. Much has been learned about the behavior of graphite in reactor environments in the 60-plus years since the first graphite rectors went into service. The extensive list of references in the Bibliography is plainly testament to this fact. Our current knowledge base is well developed. Although data are lacking for the specific grades being considered for Generation IV (Gen IV) concepts, such as the NGNP, it is fully expected that the behavior of these graphites will conform to the recognized trends for near isotropic nuclear graphite. Thus, much of the data needed is confirmatory in nature. Theories that can explain graphite behavior have been postulated and, in many cases, shown to represent experimental data well. However, these theories need to be tested against data for the new graphites and extended to higher neutron doses and temperatures pertinent to the new Gen IV reactor concepts. It is anticipated that current and planned future graphite irradiation experiments will provide the data needed to validate many of the currently accepted models, as well as providing the needed data for design confirmation.

  5. Maintaining a Technology-Neutral Approach to Hydrogen Production Process Development through Conceptual Design of the Next Generation Nuclear Plant

    SciTech Connect (OSTI)

    Michael W. Patterson

    2008-05-01T23:59:59.000Z

    The Next Generation Nuclear Plant (NGNP) project was authorized in the Energy Policy Act of 2005 (EPAct), tasking the U.S. Department of Energy (DOE) with demonstrating High Temperature Gas-Cooled Reactor (HTGR) technology. The demonstration is to include the technical, licensing, operational, and commercial viability of HTGR technology for the production of electricity and hydrogen. The Nuclear Hydrogen Initiative (NHI), a component of the DOE Hydrogen Program managed by the Office of Nuclear Energy, is also investigating multiple approaches to cost effective hydrogen production from nuclear energy. The objective of NHI is development of the technology and information basis for a future decision on commercial viability. The initiatives are clearly intertwined. While the objectives of NGNP and NHI are generally consistent, NGNP has progressed to the project definition phase and the project plan has matured. Multiple process applications for the NGNP require process heat, electricity and hydrogen in varied combinations and sizes. Coupling these processes to the reactor in multiple configurations adds complexity to the design, licensing and demonstration of both the reactor and the hydrogen production process. Commercial viability of hydrogen production may depend on the specific application and heat transport configuration. A component test facility (CTF) is planned by the NGNP to support testing and demonstration of NGNP systems, including those for hydrogen production, in multiple configurations. Engineering-scale demonstrations in the CTF are expected to start in 2012 to support scheduled design and licensing activities leading to subsequent construction and operation. Engineering-scale demonstrations planned by NHI are expected to start at least two years later. Reconciliation of these schedules is recommended to successfully complete both initiatives. Hence, closer and earlier integration of hydrogen process development and heat transport systems is sensible. For integration purposes, an analysis comparing the design, cost and schedule impact of maintaining a technology neutral approach through conceptual design or making an early hydrogen process technology selection was performed. Early selection does not specifically eliminate a technology, but rather selects the first hydrogen technology for demonstration. A systems-engineering approach was taken to define decision-making criteria for selecting a hydrogen technology. The relative technical, cost and schedule risks of each approach were analyzed and risk mitigation strategies were recommended, including provisions to maintain close collaboration with the NHI. The results of these analyses are presented here.

  6. CONSTRUCTION OF NUCLEAR POWER PLANTS

    E-Print Network [OSTI]

    CONSTRUCTION OF NUCLEAR POWER PLANTS A Workshop on "NUCLEAR ENERGY RENAISSANCE" Addressing WAS DEEPLY INVOLVED IN ALMOST EVERY ASPECT OF BUILDING THE PLANTS THROUGH · Quality Assurance · Nuclear IN CONSTRUCTION OF ST. LUCIE-2 #12;LESSONS LEARNED FROM St. Lucie-2 NUCLEAR POWER PLANTS CAN BE BUILT

  7. California Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    California nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State...

  8. Pennsylvania Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    Pennsylvania nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State...

  9. Connecticut Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    Connecticut nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State...

  10. Nuclear Power Plant Design Project

    E-Print Network [OSTI]

    Nuclear Power Plant Design Project A Response to the Environmental and Economic Challenge Of Global.............................................................................................................. 4 3. Assessment of the Issues and Needs for a New Plant

  11. Next Generation Nuclear Plant Methods Technical Program Plan

    SciTech Connect (OSTI)

    Richard R. Schultz; Abderrafi M. Ougouag; David W. Nigg; Hans D. Gougar; Richard W. Johnson; William K. Terry; Chang H. Oh; Donald W. McEligot; Gary W. Johnsen; Glenn E. McCreery; Woo Y. Yoon; James W. Sterbentz; J. Steve Herring; Temitope A. Taiwo; Thomas Y. C. Wei; William D. Pointer; Won S. Yang; Michael T. Farmer; Hussein S. Khalil; Madeline A. Feltus

    2007-01-01T23:59:59.000Z

    One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope of the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.

  12. Next Generation Nuclear Plant Methods Technical Program Plan

    SciTech Connect (OSTI)

    Richard R. Schultz; Abderrafi M. Ougouag; David W. Nigg; Hans D. Gougar; Richard W. Johnson; William K. Terry; Chang H. Oh; Donald W. McEligot; Gary W. Johnsen; Glenn E. McCreery; Woo Y. Yoon; James W. Sterbentz; J. Steve Herring; Temitope A. Taiwo; Thomas Y. C. Wei; William D. Pointer; Won S. Yang; Michael T. Farmer; Hussein S. Khalil; Madeline A. Feltus

    2010-12-01T23:59:59.000Z

    One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope of the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.

  13. Next Generation Nuclear Plant Methods Technical Program Plan -- PLN-2498

    SciTech Connect (OSTI)

    Richard R. Schultz; Abderrafi M. Ougouag; David W. Nigg; Hans D. Gougar; Richard W. Johnson; William K. Terry; Chang H. Oh; Donald W. McEligot; Gary W. Johnsen; Glenn E. McCreery; Woo Y. Yoon; James W. Sterbentz; J. Steve Herring; Temitope A. Taiwo; Thomas Y. C. Wei; William D. Pointer; Won S. Yang; Michael T. Farmer; Hussein S. Khalil; Madeline A. Feltus

    2010-09-01T23:59:59.000Z

    One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope of the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.

  14. Louisiana Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    Louisiana nuclear power plants, summer capacity and net generation, 2010" "Plant NameTotal Reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

  15. Wisconsin Nuclear Profile - Point Beach Nuclear Plant

    U.S. Energy Information Administration (EIA) Indexed Site

    Point Beach Nuclear Plant" "Unit","Summer capacity (mw)","Net generation (thousand mwh)","Summer capacity factor (percent)","Type","Commercial operation date","License expiration...

  16. Tennessee Nuclear Profile - Watts Bar Nuclear Plant

    U.S. Energy Information Administration (EIA) Indexed Site

    Watts Bar Nuclear Plant" "Unit","Summer capacity (mw)","Net generation (thousand mwh)","Summer capacity factor (percent)","Type","Commercial operation date","License expiration...

  17. NGNP Graphite Testing and Qualification Specimen Selection Strategy

    SciTech Connect (OSTI)

    Robert Bratton

    2005-05-01T23:59:59.000Z

    The available grades of graphite for the NGNP are reviewed. A selection matrix is presented outlining the available grades for the NGNP graphite irradiation program based upon input from potential NGNP vendors, graphite manufactures, and graphite experts.

  18. Status of the NGNP graphite creep experiments AGC-1 and AGC-2 irradiated in the advanced test reactor

    SciTech Connect (OSTI)

    S. Blaine Grover

    2014-05-01T23:59:59.000Z

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the next generation nuclear plant (NGNP) very high temperature gas reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six peripheral stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six peripheral stacks will have three different compressive loads applied to the top half of three diametrically opposite pairs of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during irradiation of the experiment.

  19. Kansas Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    nuclear net generation (percent)","Owner" "Wolf Creek Generating Station Unit 1","1,160","9,556",100.0,"Wolf Creek Nuclear Optg Corp" "1 Plant 1 Reactor","1,160","9,556",100.0...

  20. Vermont Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    mwh)","Share of State nuclear net generation (percent)","Owner" "Vermont Yankee Unit 1",620,"4,782",100.0,"Entergy Nuclear Vermont Yankee" "1 Plant 1 Reactor",620,"4,782",100.0...

  1. Massachusetts Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    (percent)","Owner" "Pilgrim Nuclear Power Station Unit 1",685,"5,918",100.0,"Entergy Nuclear Generation Co" "1 Plant 1 Reactor",685,"5,918",100.0 "Note: Totals may not equal...

  2. Issues for New Nuclear Plants

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    to Explore * Idaho's energy picture * Nuclear power in the U.S. * Potential for a nuclear power plant in Idaho 0 5 10 15 20 25 1960 1970 1980 1990 2000 Million Megawatt-Hours Total...

  3. Issues for New Nuclear Plants

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    to produce heavy components and nuclear-grade equipment - Transportation of heavy components - Constructionoperation workforce - Cost of new plants Cooling Technology...

  4. Dependable Hydrogen and Industrial Heat Generation from the Next Generation Nuclear Plant

    SciTech Connect (OSTI)

    Charles V. Park; Michael W. Patterson; Vincent C. Maio; Piyush Sabharwall

    2009-03-01T23:59:59.000Z

    The Department of Energy is working with industry to develop a next generation, high-temperature gas-cooled nuclear reactor (HTGR) as a part of the effort to supply the US with abundant, clean and secure energy. The Next Generation Nuclear Plant (NGNP) project, led by the Idaho National Laboratory, will demonstrate the ability of the HTGR to generate hydrogen, electricity, and high-quality process heat for a wide range of industrial applications. Substituting HTGR power for traditional fossil fuel resources reduces the cost and supply vulnerability of natural gas and oil, and reduces or eliminates greenhouse gas emissions. As authorized by the Energy Policy Act of 2005, industry leaders are developing designs for the construction of a commercial prototype producing up to 600 MWt of power by 2021. This paper describes a variety of critical applications that are appropriate for the HTGR with an emphasis placed on applications requiring a clean and reliable source of hydrogen. An overview of the NGNP project status and its significant technology development efforts are also presented.

  5. Owners of nuclear power plants

    SciTech Connect (OSTI)

    Hudson, C.R.; White, V.S.

    1996-11-01T23:59:59.000Z

    Commercial nuclear power plants in this country can be owned by a number of separate entities, each with varying ownership proportions. Each of these owners may, in turn, have a parent/subsidiary relationship to other companies. In addition, the operator of the plant may be a different entity as well. This report provides a compilation on the owners/operators for all commercial power reactors in the United States. While the utility industry is currently experiencing changes in organizational structure which may affect nuclear plant ownership, the data in this report is current as of July 1996. The report is divided into sections representing different aspects of nuclear plant ownership.

  6. North Carolina Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    Carolina nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

  7. New Jersey Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

  8. New York Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

  9. Heat exchanger temperature response for duty-cycle transients in the NGNP/HTE.

    SciTech Connect (OSTI)

    Vilim, R. B.; Nuclear Engineering Division

    2009-03-12T23:59:59.000Z

    Control system studies were performed for the Next Generation Nuclear Plant (NGNP) interfaced to the High Temperature Electrolysis (HTE) plant. Temperature change and associated thermal stresses are important factors in determining plant lifetime. In the NGNP the design objective of a 40 year lifetime for the Intermediate Heat Exchanger (IHX) in particular is seen as a challenge. A control system was designed to minimize temperature changes in the IHX and more generally at all high-temperature locations in the plant for duty-cycle transients. In the NGNP this includes structures at the reactor outlet and at the inlet to the turbine. This problem was approached by identifying those high-level factors that determine temperature rates of change. First are the set of duty cycle transients over which the control engineer has little control but which none-the-less must be addressed. Second is the partitioning of the temperature response into a quasi-static component and a transient component. These two components are largely independent of each other and when addressed as such greater understanding of temperature change mechanisms and how to deal with them is achieved. Third is the manner in which energy and mass flow rates are managed. Generally one aims for a temperature distribution that minimizes spatial non-uniformity of thermal expansion in a component with time. This is can be achieved by maintaining a fixed spatial temperature distribution in a component during transients. A general rule of thumb for heat exchangers is to maintain flow rate proportional to thermal power. Additionally the product of instantaneous flow rate and heat capacity should be maintained the same on both sides of the heat exchanger. Fourth inherent mechanisms for stable behavior should not be compromised by active controllers that can introduce new feedback paths and potentially create under-damped response. Applications of these principles to the development of a plant control strategy for the reference NGNP/HTE plant can be found in the body of this report. The outcome is an integrated plant/control system design. The following conclusions are drawn from the analysis: (1) The plant load schedule can be managed to maintain near-constant hot side temperatures over the load range in both the nuclear and chemical plant. (2) The reactor open-loop response is inherently stable resulting mainly from a large Doppler temperature coefficient compared to the other reactivity temperature feedbacks. (3) The typical controller used to manage reactor power production to maintain reactor outlet temperature at a setpoint introduces a feedback path that tends to destabilize reactor power production in the NGNP. (4) A primary loop flow controller that forces primary flow to track PCU flow rate is effective in minimizing spatial temperature differentials within the IHX. (5) Inventory control in both the primary and PCU system during ramp load change transients is an effective means of maintaining high NGNP thermal efficiency while at reduced electric load. (6) Turbine bypass control is an effective means for responding to step changes in generator load when equipment capacity limitations prevent inventory control from being effective. (7) Turbine bypass control is effective in limiting PCU shaft over speed for the loss of generator load upset event. (8) The proposed control strategy is effective in limiting time variation of the differential spatial temperature distribution in the IHX during transients. Essentially the IHX can be made to behave in a manner where each point in the IHX experiences approximately the same temperature rate of change during a transient. (9) The stability of the closed-loop Brayton cycle was found to be sensitive to where one operates on the turbo-machine performance maps. There are competing interests: more stable operation means operating on the curves at points that reduce overall cycle efficiency. Future work should address in greater detail elements that came to light in the course of this work. Specifically: (1) A stability analysi

  10. Next Generation Nuclear Plant Defense-in-Depth Approach

    SciTech Connect (OSTI)

    Edward G. Wallace; Karl N. Fleming; Edward M. Burns

    2009-12-01T23:59:59.000Z

    The purpose of this paper is to (1) document the definition of defense-in-depth and the pproach that will be used to assure that its principles are satisfied for the NGNP project and (2) identify the specific questions proposed for preapplication discussions with the NRC. Defense-in-depth is a safety philosophy in which multiple lines of defense and conservative design and evaluation methods are applied to assure the safety of the public. The philosophy is also intended to deliver a design that is tolerant to uncertainties in knowledge of plant behavior, component reliability or operator performance that might compromise safety. This paper includes a review of the regulatory foundation for defense-in-depth, a definition of defense-in-depth that is appropriate for advanced reactor designs based on High Temperature Gas-cooled Reactor (HTGR) technology, and an explanation of how this safety philosophy is achieved in the NGNP.

  11. Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 1: Main Report

    SciTech Connect (OSTI)

    Ball, Sydney J [ORNL

    2008-03-01T23:59:59.000Z

    A phenomena identification and ranking table (PIRT) process was conducted for the Next Generation Nuclear Plant (NGNP) design. This design (in the conceptual stage) is a modular high-temperature gas-cooled reactor (HTGR) that generates both electricity and process heat for hydrogen production. Expert panels identified safety-relevant phenomena, ranked their importance, and assessed the knowledge levels in the areas of accidents and thermal fluids, fission-product transport and dose, high-temperature materials, graphite, and process heat for hydrogen production. This main report summarizes and documents the process and scope of the reviews, noting the major activities and conclusions. The identified phenomena, analyses, rationales, and associated ratings of the phenomena, plus a summary of each panel's findings, are presented. Individual panel reports for these areas are provided as attached volumes to this main report and provide considerably more detail about each panel's deliberations as well as a more complete listing of the phenomena that were evaluated.

  12. Microsoft Word - NGNP_Contamination Control Report-050108 FINAL...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    94 NGNP-NHS 50-CC Contamination Control 8 of 101 NGNP-Contamination Control Report-050108.doc 2008 Westinghouse Electric Company LLC LIST OF FIGURES Figure 1-1...

  13. Status of the NGNP Graphite Creep Experiments AGC-1 and AGC-2 Irradiated in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Blaine Grover

    2012-10-01T23:59:59.000Z

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six peripheral stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six peripheral stacks will have different compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during irradiation of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. This paper will briefly discuss the design of the experiment and control systems, and then present the irradiation results for each experiment to date.

  14. South Carolina Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    South Carolina nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State...

  15. Next Generation Nuclear Plant Methods Research and Development Technical Program Plan -- PLN-2498

    SciTech Connect (OSTI)

    Richard R. Schultz; Abderrafi M. Ougouag; David W. Nigg; Hans D. Gougar; Richard W. Johnson; William K. Terry; Chang H. Oh; Donald W. McEligot; Gary W. Johnsen; Glenn E. McCreery; Woo Y. Yoon; James W. Sterbentz; J. Steve Herring; Temitope A. Taiwo; Thomas Y. C. Wei; William D. Pointer; Won S. Yang; Michael T. Farmer; Hussein S. Khalil; Madeline A. Feltus

    2008-09-01T23:59:59.000Z

    One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope of the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.

  16. Nuclear power plants: structure and function

    SciTech Connect (OSTI)

    Hendrie, J.M.

    1983-01-01T23:59:59.000Z

    Topics discussed include: steam electric plants; BWR type reactors; PWR type reactors; thermal efficiency of light water reactors; other types of nuclear power plants; the fission process and nuclear fuel; fission products and reactor afterheat; and reactor safety.

  17. Advanced nuclear plant control complex

    DOE Patents [OSTI]

    Scarola, Kenneth (Windsor, CT); Jamison, David S. (Windsor, CT); Manazir, Richard M. (North Canton, CT); Rescorl, Robert L. (Vernon, CT); Harmon, Daryl L. (Enfield, CT)

    1993-01-01T23:59:59.000Z

    An advanced control room complex for a nuclear power plant, including a discrete indicator and alarm system (72) which is nuclear qualified for rapid response to changes in plant parameters and a component control system (64) which together provide a discrete monitoring and control capability at a panel (14-22, 26, 28) in the control room (10). A separate data processing system (70), which need not be nuclear qualified, provides integrated and overview information to the control room and to each panel, through CRTs (84) and a large, overhead integrated process status overview board (24). The discrete indicator and alarm system (72) and the data processing system (70) receive inputs from common plant sensors and validate the sensor outputs to arrive at a representative value of the parameter for use by the operator during both normal and accident conditions, thereby avoiding the need for him to assimilate data from each sensor individually. The integrated process status board (24) is at the apex of an information hierarchy that extends through four levels and provides access at each panel to the full display hierarchy. The control room panels are preferably of a modular construction, permitting the definition of inputs and outputs, the man machine interface, and the plant specific algorithms, to proceed in parallel with the fabrication of the panels, the installation of the equipment and the generic testing thereof.

  18. Plant nuclear bodies Peter J Shaw1

    E-Print Network [OSTI]

    Shaw, Peter

    Plant nuclear bodies Peter J Shaw1 and John WS Brown2 Knowledge of the organization bodies have been examined in plants, and recently, various other sub-nuclear domains that are involved. Until recently, the only plant nuclear bodies to be in any way characterized were the nucleolus [11

  19. NGNP Point Design - Results of the Initial Neutronics and Thermal-Hydraulic Assessments During FY-03, Rev. 1

    SciTech Connect (OSTI)

    Philip E. MacDonald; James W. Sterbentz; Robert L. Sant; P. Bayless; H. D. Gougar; R. L. Moore; A. M. Ougouag; W. K. Terry

    2003-09-01T23:59:59.000Z

    This report presents the preliminary preconceptual designs for two possible versions of the Next Generation Nuclear Plant (NGNP), one for a prismatic fuel type helium gas-cooled reactor and one for a pebble bed fuel helium gas reactor. Both designs are to meet three basic requirements: a coolant outlet temperature of 1000 °C, passive safety, and a total power output consistent with that expected for commercial high-temperature gas-cooled reactors. The two efforts are discussed separately below. The analytical results presented in this report are very promising, however, we wish to caution the reader that future, more detailed, design work will be needed to provide final answers to a number of key questions including the allowable power level, the inlet temperature, the power density, the optimum fuel form, and others. The point design work presented in this report provides a starting point for other evaluations, and directions for the detailed design, but not final answers.

  20. NGNP Composites R&D Technical Issues Study

    SciTech Connect (OSTI)

    AREVA Federal Services

    2008-09-01T23:59:59.000Z

    This study identifies potential applications and design requirements for ceramic materials (CMs) and ceramic composite materials (CCMs) in the NGNP hightemperature gas-cooled reactor (HTGR) primary circuit. Components anticipated for fabrication from non-graphite CMs and CCMs are identified along with recommended normal and off-normal operating conditions. The evaluation defines required dimensions and material properties of the candidate materials for normal operating conditions (NOC), anticipated transients, abnormal events, and design basis events. The report also identifies additional activities required for codifying the selected materials. The activities include ASTM Standard and ASME Code development and other work to support NRC licensing of the plant. Evaluation of the NGNP baseline design indicates components requiring either CMs or CCMs depend upon the reactor operating temperatures. For a reactor outlet temperature of 900 oC, four of the five evaluated components would benefit from either CMs or CCMs. Although some thermal and mechanical data exist for most of the candidate materials, they all need additional irradiation, thermal, and mechanical testing. The codification process must take into account the type of material and the geometry of components using either CMs or CCMs. The process requires close integration of the design and the research and development (R&D) program, which has already started by using preliminary control rod component designs as the basis for establishing specimen geometry and test conditions. The remaining time and budget for completing the R&D program need further assessment.

  1. Sabotage at Nuclear Power Plants

    SciTech Connect (OSTI)

    Purvis, James W.

    1999-07-21T23:59:59.000Z

    Recently there has been a noted worldwide increase in violent actions including attempted sabotage at nuclear power plants. Several organizations, such as the International Atomic Energy Agency and the US Nuclear Regulatory Commission, have guidelines, recommendations, and formal threat- and risk-assessment processes for the protection of nuclear assets. Other examples are the former Defense Special Weapons Agency, which used a risk-assessment model to evaluate force-protection security requirements for terrorist incidents at DOD military bases. The US DOE uses a graded approach to protect its assets based on risk and vulnerability assessments. The Federal Aviation Administration and Federal Bureau of Investigation conduct joint threat and vulnerability assessments on high-risk US airports. Several private companies under contract to government agencies use formal risk-assessment models and methods to identify security requirements. The purpose of this paper is to survey these methods and present an overview of all potential types of sabotage at nuclear power plants. The paper discusses emerging threats and current methods of choice for sabotage--especially vehicle bombs and chemical attacks. Potential consequences of sabotage acts, including economic and political; not just those that may result in unacceptable radiological exposure to the public, are also discussed. Applicability of risk-assessment methods and mitigation techniques are also presented.

  2. The Politically Correct Nuclear Energy Plant

    E-Print Network [OSTI]

    The Politically Correct Nuclear Energy Plant Andrew C. Kadak Massachusetts Institute of Technology - Small is Beautiful · Nuclear Energy - But Getting Better #12;Politically Correct ! · Natural Safety is a bad idea. · There is no new nuclear energy plant that is competitive at this time. · De-regulation did

  3. Microsoft Word - NGNP-Bounding Reqts for Project Development...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    under normal and accident conditions, tritium generation, and characterizing the permeability of the materials of construction anticipated for use in the NGNP heat exchange...

  4. KEY DESIGN REQUIREMENTS FOR THE HIGH TEMPERATURE GAS-COOLED REACTOR NUCLEAR HEAT SUPPLY SYSTEM

    SciTech Connect (OSTI)

    L.E. Demick

    2010-09-01T23:59:59.000Z

    Key requirements that affect the design of the high temperature gas-cooled reactor nuclear heat supply system (HTGR-NHSS) as the NGNP Project progresses through the design, licensing, construction and testing of the first of a kind HTGR based plant are summarized. These requirements derive from pre-conceptual design development completed to-date by HTGR Suppliers, collaboration with potential end users of the HTGR technology to identify energy needs, evaluation of integration of the HTGR technology with industrial processes and recommendations of the NGNP Project Senior Advisory Group.

  5. Nuclear Plant/Hydrogen Plant Safety: Issues and Approaches

    SciTech Connect (OSTI)

    Steven R. Sherman

    2007-06-01T23:59:59.000Z

    The U.S. Department of Energy, through its agents the Next Generation Nuclear Plant Project and the Nuclear Hydrogen Initiative, is working on developing the technologies to enable the large scale production of hydrogen using nuclear power. A very important consideration in the design of a co-located and connected nuclear plant/hydrogen plant facility is safety. This study provides an overview of the safety issues associated with a combined plant and discusses approaches for categorizing, quantifying, and addressing the safety risks.

  6. NGNP Program 2013 Status and Path Forward

    SciTech Connect (OSTI)

    Hans Gougar

    2014-03-01T23:59:59.000Z

    High temperature gas-cooled reactor (HTGR) technology can play an important role in the energy future of the United States by extending the use of nuclear energy for non-electricity energy production missions, as well as continuing to provide a considerable base load electric power generation capability. Extending nuclear energy into the industrial and transportation sectors through the coproduction of process heat and electricity provides safe, reliable energy for these sectors in an environmentally responsible manner. The modular HTGR provides a substantial improvement in nuclear plant safety for the protection of the public and the environment, and supports collocation of the HTGRhigh temperature gas-cooled reactor with major industrial facilities. Under U.S. Department of Energy direction since 2006, the Next Generation Nuclear Plant Project at Idaho National Laboratory has been working toward commercializing the HTGR technology. However, a recent decision by the Secretary of Energy to reduce the scope of the Next Generation Nuclear Plant Project to a research and development program, considerable realignment has taken place. This report: (1) summarizes the accomplishments of the Next Generation Nuclear Plant Program from FY2011 through FY2013; (2) lays out the path forward necessary to achieve the ultimate objective of commercializing HTGR technology; and (3) discusses ongoing technical, licensing, and evaluation activities under the realigned Next Generation Nuclear Plant program considered important to preserve the significant investment made by the government to-date and to maintain some progress in meeting the objectives of the Energy Policy Act of 2005 (EPAct2005).

  7. Maryland Nuclear Profile - Calvert Cliffs Nuclear Power Plant

    U.S. Energy Information Administration (EIA) Indexed Site

    Calvert Cliffs Nuclear Power Plant" "Unit","Summer capacity (mw)","Net generation (thousand mwh)","Summer capacity factor (percent)","Type","Commercial operation date","License...

  8. New York Nuclear Profile - R E Ginna Nuclear Power Plant

    U.S. Energy Information Administration (EIA) Indexed Site

    R E Ginna Nuclear Power Plant" "Unit","Summer Capacity (MW)","Net Generation (Thousand MWh)","Summer Capacity Factor (Percent)","Type","Commercial Operation Date","License...

  9. Future AI and Robotics Technology for Nuclear Plants Decommissioning

    E-Print Network [OSTI]

    Hu, Huosheng

    Future AI and Robotics Technology for Nuclear Plants Decommissioning Huosheng Hu and Liam Cragg to aid in decommissioning nuclear plants that have been used to process or store nuclear materials. Scope potential applications to nuclear plant decommissioning, namely Nanotechnology, Telepresence

  10. Organizational learning at nuclear power plants

    E-Print Network [OSTI]

    Carroll, John S.

    1991-01-01T23:59:59.000Z

    The Nuclear Power Plant Advisory Panel on Organizational Learning provides channels of communications between the management and organization research projects of the MIT International Program for Enhanced Nuclear Power ...

  11. The second and third NGNP advanced gas reactor fuel irradiation experiments

    SciTech Connect (OSTI)

    Grover, S. B.; Petti, D. A. [Idaho National Laboratory, 2525 N. Fremont Ave., Idaho Falls, ID 83415 (United States)

    2012-07-01T23:59:59.000Z

    The United States Dept. of Energy's Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is currently scheduled to irradiate a total of five low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The irradiations are being accomplished to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas cooled reactors. The experiments will each consist of at least six separate capsules, and will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The effluent sweep gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and completed a very successful irradiation in early November 2009. The second experiment (AGR-2) started irradiation in June 2010, and the third and fourth experiments have been combined into a single larger irradiation (AGR-3/4) that is currently being assembled. The design and status of the second through fourth experiments as well as the irradiation results of the second experiment to date are discussed. (authors)

  12. High Temperature Gas-Cooled Reactors Lessons Learned Applicable to the Next Generation Nuclear Plant

    SciTech Connect (OSTI)

    J. M. Beck; L. F. Pincock

    2011-04-01T23:59:59.000Z

    The purpose of this report is to identify possible issues highlighted by these lessons learned that could apply to the NGNP in reducing technical risks commensurate with the current phase of design. Some of the lessons learned have been applied to the NGNP and documented in the Preconceptual Design Report. These are addressed in the background section of this document and include, for example, the decision to use TRISO fuel rather than BISO fuel used in the Peach Bottom reactor; the use of a reactor pressure vessel rather than prestressed concrete found in Fort St. Vrain; and the use of helium as a primary coolant rather than CO2. Other lessons learned, 68 in total, are documented in Sections 2 through 6 and will be applied, as appropriate, in advancing phases of design. The lessons learned are derived from both negative and positive outcomes from prior HTGR experiences. Lessons learned are grouped according to the plant, areas, systems, subsystems, and components defined in the NGNP Preconceptual Design Report, and subsequent NGNP project documents.

  13. Multi-Scale Multi-physics Methods Development for the Calculation of Hot-Spots in the NGNP

    SciTech Connect (OSTI)

    Downar, Thomas; Seker, Volkan

    2013-04-30T23:59:59.000Z

    Radioactive gaseous fission products are released out of the fuel element at a significantly higher rate when the fuel temperature exceeds 1600°C in high-temperature gas-cooled reactors (HTGRs). Therefore, it is of paramount importance to accurately predict the peak fuel temperature during all operational and design-basis accident conditions. The current methods used to predict the peak fuel temperature in HTGRs, such as the Next-Generation Nuclear Plant (NGNP), estimate the average fuel temperature in a computational mesh modeling hundreds of fuel pebbles or a fuel assembly in a pebble-bed reactor (PBR) or prismatic block type reactor (PMR), respectively. Experiments conducted in operating HTGRs indicate considerable uncertainty in the current methods and correlations used to predict actual temperatures. The objective of this project is to improve the accuracy in the prediction of local "hot" spots by developing multi-scale, multi- physics methods and implementing them within the framework of established codes used for NGNP analysis. The multi-scale approach which this project will implement begins with defining suitable scales for a physical and mathematical model and then deriving and applying the appropriate boundary conditions between scales. The macro scale is the greatest length that describes the entire reactor, whereas the meso scale models only a fuel block in a prismatic reactor and ten to hundreds of pebbles in a pebble bed reactor. The smallest scale is the micro scale--the level of a fuel kernel of the pebble in a PBR and fuel compact in a PMR--which needs to be resolved in order to calculate the peak temperature in a fuel kernel.

  14. Standard Problems for CFD Validation for NGNP - Status Report

    SciTech Connect (OSTI)

    Richard W. Johnson; Richard R. Schultz

    2010-08-01T23:59:59.000Z

    The U.S. Department of Energy (DOE) is conducting research and development to support the resurgence of nuclear power in the United States for both electrical power generation and production of process heat required for industrial processes such as the manufacture of hydrogen for use as a fuel in automobiles. The project is called the Next Generation Nuclear Plant (NGNP) Project, which is based on a Generation IV reactor concept called the very high temperature reactor (VHTR). The VHTR will be of the prismatic or pebble bed type; the former is considered herein. The VHTR will use helium as the coolant at temperatures ranging from 250°C to perhaps 1000°C. While computational fluid dynamics (CFD) has not previously been used for the safety analysis of nuclear reactors in the United States, it is being considered for existing and future reactors. It is fully recognized that CFD simulation codes will have to be validated for flow physics reasonably close to actual fluid dynamic conditions expected in normal operational and accident situations. The “Standard Problem” is an experimental data set that represents an important physical phenomenon or phenomena, whose selection is based on a phenomena identification and ranking table (PIRT) for the reactor in question. It will be necessary to build a database that contains a number of standard problems for use to validate CFD and systems analysis codes for the many physical problems that will need to be analyzed. The first two standard problems that have been developed for CFD validation consider flow in the lower plenum of the VHTR and bypass flow in the prismatic core. Both involve scaled models built from quartz and designed to be installed in the INL’s matched index of refraction (MIR) test facility. The MIR facility employs mineral oil as the working fluid at a constant temperature. At this temperature, the index of refraction of the mineral oil is the same as that of the quartz. This provides an advantage to the optics used for data gathering. Particle image velocimetry (PIV) is used to take the data. The first standard problem represents several flow physics expected to be present in the lower plenum of the prismatic VHTR. In the lower plenum, heated helium coolant in the form of jets issues downward into the plenum and is then forced to turn ninety degrees and flow toward the exit duct. The lower plenum is filled with cylindrical graphite posts that hold up the core. Figure S-1 provides a plan view of the lower plenum. The red circles represent support posts holding up columns of heated blocks. Grey circles represent support posts under columns of reflector blocks. Helium enters the lower plenum at the junctions of the hexagonal blocks.

  15. Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 4: High-Temperature Materials PIRTs

    SciTech Connect (OSTI)

    Corwin, William R [ORNL; Ballinger, R. [Massachusetts Institute of Technology (MIT); Majumdar, S. [Argonne National Laboratory (ANL); Weaver, K. D. [Idaho National Laboratory (INL)

    2008-03-01T23:59:59.000Z

    The Phenomena Identification and Ranking Table (PIRT) technique was used to identify safety-relevant/safety-significant phenomena and assess the importance and related knowledge base of high-temperature structural materials issues for the Next Generation Nuclear Plant (NGNP), a very high temperature gas-cooled reactor (VHTR). The major aspects of materials degradation phenomena that may give rise to regulatory safety concern for the NGNP were evaluated for major structural components and the materials comprising them, including metallic and nonmetallic materials for control rods, other reactor internals, and primary circuit components; metallic alloys for very high-temperature service for heat exchangers and turbomachinery, metallic alloys for high-temperature service for the reactor pressure vessel (RPV), other pressure vessels and components in the primary and secondary circuits; and metallic alloys for secondary heat transfer circuits and the balance of plant. These materials phenomena were primarily evaluated with regard to their potential for contributing to fission product release at the site boundary under a variety of event scenarios covering normal operation, anticipated transients, and accidents. Of all the high-temperature metallic components, the one most likely to be heavily challenged in the NGNP will be the intermediate heat exchanger (IHX). Its thin, internal sections must be able to withstand the stresses associated with thermal loading and pressure drops between the primary and secondary loops under the environments and temperatures of interest. Several important materials-related phenomena related to the IHX were identified, including crack initiation and propagation; the lack of experience of primary boundary design methodology limitations for new IHX structures; and manufacturing phenomena for new designs. Specific issues were also identified for RPVs that will likely be too large for shop fabrication and transportation. Validated procedures for on-site welding, post-weld heat treatment (PWHT), and inspections will be required for the materials of construction. High-importance phenomena related to the RPV include crack initiation and subcritical crack growth; field fabrication process control; property control in heavy sections; and the maintenance of high emissivity of the RPV materials over their service lifetime to enable passive heat rejection from the reactor core. All identified phenomena related to the materials of construction for the IHX, RPV, and other components were evaluated and ranked for their potential impact on reactor safety.

  16. Status of the Combined Third and Fourth NGNP Fuel Irradiations In the Advanced Test Reactor

    SciTech Connect (OSTI)

    S. Blaine Grover; David A. Petti; Michael E. Davenport

    2013-07-01T23:59:59.000Z

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in September 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. Since the purpose of this combined experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is significantly different from the first two experiments, though the control and monitoring systems are extremely similar. The design of the experiment will be discussed followed by its progress and status to date.

  17. Next Generation Nuclear Plant Phenomena

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Components," Journal of Nuclear Materials, 212-215, 1223 (1994). 13. Arnold, L, Windscale 1957, Anatomy of a Nuclear Accident, St Martin Press, London, 1992. 14....

  18. Nuclear Power Plant Concrete Structures

    SciTech Connect (OSTI)

    Basu, Prabir [International Atomic Energy Agency (IAEA)] [International Atomic Energy Agency (IAEA); Labbe, Pierre [Electricity of France (EDF)] [Electricity of France (EDF); Naus, Dan [Oak Ridge National Laboratory (ORNL)] [Oak Ridge National Laboratory (ORNL)

    2013-01-01T23:59:59.000Z

    A nuclear power plant (NPP) involves complex engineering structures that are significant items of the structures, systems and components (SSC) important to the safe and reliable operation of the NPP. Concrete is the commonly used civil engineering construction material in the nuclear industry because of a number of advantageous properties. The NPP concrete structures underwent a great degree of evolution, since the commissioning of first NPP in early 1960. The increasing concern with time related to safety of the public and environment, and degradation of concrete structures due to ageing related phenomena are the driving forces for such evolution. The concrete technology underwent rapid development with the advent of chemical admixtures of plasticizer/super plasticizer category as well as viscosity modifiers and mineral admixtures like fly ash and silica fume. Application of high performance concrete (HPC) developed with chemical and mineral admixtures has been witnessed in the construction of NPP structures. Along with the beneficial effect, the use of admixtures in concrete has posed a number of challenges as well in design and construction. This along with the prospect of continuing operation beyond design life, especially after 60 years, the impact of extreme natural events ( as in the case of Fukushima NPP accident) and human induced events (e.g. commercial aircraft crash like the event of September 11th 2001) has led to further development in the area of NPP concrete structures. The present paper aims at providing an account of evolution of NPP concrete structures in last two decades by summarizing the development in the areas of concrete technology, design methodology and construction techniques, maintenance and ageing management of concrete structures.

  19. aguirre nuclear plant: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Denmark December 1991 12;Abstract. A computer model of a simplified pressurized nuclear power plant a compute simulation of a simplified pressurized nuclear power plant model...

  20. DOE Announces Loan Guarantee Applications for Nuclear Power Plant...

    Energy Savers [EERE]

    Loan Guarantee Applications for Nuclear Power Plant Construction DOE Announces Loan Guarantee Applications for Nuclear Power Plant Construction October 2, 2008 - 3:43pm Addthis...

  1. RESEARCH PAPER Composition of the plant nuclear envelope: theme and

    E-Print Network [OSTI]

    Meier, Iris

    RESEARCH PAPER Composition of the plant nuclear envelope: theme and variations Iris Meier* Plant plants is only just beginning, fundamental differences from the animal nuclear envelope have already been to known plant regulatory pathways. Plant nuclear envelope composition The inner nuclear envelope A number

  2. NGNP/HTE full-power operation at reduced high-temperature heat exchanger temperatures.

    SciTech Connect (OSTI)

    VIlim, R.; Nuclear Engineering Division

    2009-03-12T23:59:59.000Z

    Operation of the Next Generation Nuclear Plant (NGNP) with reduced reactor outlet temperature at full power was investigated for the High Temperature Electrolysis (HTE) hydrogen-production application. The foremost challenge for operation at design temperature is achieving an acceptably long service life for heat exchangers. In both the Intermediate Heat Exchanger (IHX) and the Process Heat Exchanger (PHX) (referred to collectively as high temperature heat exchangers) a pressure differential of several MPa exists with temperatures at or above 850 C. Thermal creep of the heat exchanger channel wall may severely limit heat exchanger life depending on the alloy selected. This report investigates plant performance with IHX temperatures reduced by lowering reactor outlet temperature. The objective is to lower the temperature in heat transfer channels to the point where existing materials can meet the 40 year lifetime needed for this component. A conservative estimate for this temperature is believed to be about 700 C. The reactor outlet temperature was reduced from 850 C to 700 C while maintaining reactor power at 600 MWt and high pressure compressor outlet at 7 MPa. We included a previously reported design option for reducing temperature at the PHX. Heat exchanger lengths were adjusted to reflect the change in performance resulting from coolant property changes and from resizing related to operating-point change. Turbomachine parameters were also optimized for the new operating condition. An integrated optimization of the complete system including heat transfer equipment was not performed. It is estimated, however, that by performing a pinch analysis the combined plant efficiency can be increased from 35.5 percent obtained in this report to a value between 38.5 and 40.1 percent. Then after normalizing for a more than three percent decrease in commodities inventory compared to the reference plant, the commodities-normalized efficiency lies between 40.0 and 41.3. This compares with a value of 43.9 for the reference plant. This latter plant has a reactor outlet temperature of 850 C and the two high temperature heat exchangers. The reduction in reactor outlet temperature from 850 C to 700 C reduces the tritium permeability rate in the IHX metal by a factor of three and thermal creep by five orders of magnitude. The design option for reducing PHX temperature from 800 C to 200 C reduces the permeability there by three orders of magnitude. In that design option this heat exchanger is the single 'choke-point' for tritium migration from the nuclear to the chemical plant.

  3. Washington Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    total reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net generation (percent)","Owner" "Columbia Generating Station Unit...

  4. Iowa Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    total reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net generation (percent)","Owner" "Duane Arnold Energy Center Unit...

  5. CCN 217493 - Provide a Progress Report Describing the Status...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    1 of 6 Next Generation Nuclear Plant Licensing NGNP FUEL QUALIFICATION WHITE PAPER DEVELOPMENT STATUS REPORT August 2009 Introduction The Next Generation Nuclear Plant (NGNP)...

  6. Dose reduction at nuclear power plants

    SciTech Connect (OSTI)

    Baum, J.W.; Dionne, B.J.

    1983-01-01T23:59:59.000Z

    The collective dose equivalent at nuclear power plants increased from 1250 rem in 1969 to nearly 54,000 rem in 1980. This rise is attributable primarily to an increase in nuclear generated power from 1289 MW-y to 29,155 MW-y; and secondly, to increased average plant age. However, considerable variation in exposure occurs from plant to plant depending on plant type, refueling, maintenance, etc. In order to understand the factors influencing these differences, an investigation was initiated to study dose-reduction techniques and effectiveness of as low as reasonably achievable (ALARA) planning at light water plants. Objectives are to: identify high-dose maintenance tasks and related dose-reduction techniques; investigate utilization of high-reliability, low-maintenance equipment; recommend improved radioactive waste handling equipment and procedures; examine incentives for dose reduction; and compile an ALARA handbook.

  7. Sun-Sentinel How Florida's nuclear plants compare to Japan's

    E-Print Network [OSTI]

    Belogay, Eugene A.

    Sun-Sentinel How Florida's nuclear plants compare to Japan's By Julie Patel March 17, 2011 01:35 PM What went wrong at the Fukushima nuclear plant in Japan and how are Florida's nuclear plants prepared to deal with similar problems? Nuclear operators in Florida say the biggest risk their plants face is from

  8. NGNP High Temperature Materials White Paper

    SciTech Connect (OSTI)

    Lew Lommers; George Honma

    2012-08-01T23:59:59.000Z

    This white paper is one in a series of white papers that address key generic issues of the combined construction and operating license (COL) pre-application program key generic issues for the Next Generation Nuclear Plant reactor using the prismatic block fuel technology. The purpose of the pre-application program interactions with the NRC staff is to reduce the time required for COL application review by identifying and addressing key regulatory issues and, if possible, obtaining agreements for their resolution

  9. Missouri Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    total reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net generation (percent)","Owner" "Callaway Unit 1","1,190","8,996",100.0,"Union...

  10. Mississippi Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    total reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net generation (percent)","Owner" "Grand Gulf Unit 1","1,251","9,643",100.0,"Syste...

  11. Next Generation Nuclear Plant Phenomena

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Should that prove to be impractical (e.g. due to excessive heat loss in the intermediate heat transfer loop), an earthen berm separating the two plants may be a suitable...

  12. Status of the NGNP fuel experiment AGR-2 irradiated in the advanced test reactor

    SciTech Connect (OSTI)

    S. Blaine Grover; David A. Petti

    2014-05-01T23:59:59.000Z

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also undergo on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and sup

  13. Status of the NGNP Fuel Experiment AGR-2 Irradiated in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Blaine Grover

    2012-10-01T23:59:59.000Z

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and support systems will be briefly discussed, followed by the progress and status of the experiment to date.

  14. Nuclear plant cancellations: causes, costs, and consequences

    SciTech Connect (OSTI)

    Not Available

    1983-04-01T23:59:59.000Z

    This study was commissioned in order to help quantify the effects of nuclear plant cancellations on the Nation's electricity prices. This report presents a historical overview of nuclear plant cancellations through 1982, the costs associated with those cancellations, and the reasons that the projects were terminated. A survey is presented of the precedents for regulatory treatment of the costs, the specific methods of cost recovery that were adopted, and the impacts of these decisions upon ratepayers, utility stockholders, and taxpayers. Finally, the report identifies a series of other nuclear plants that remain at risk of canellation in the future, principally as a result of similar demand, finance, or regulatory problems cited as causes of cancellation in the past. The costs associated with these potential cancellations are estimated, along with their regional distributions, and likely methods of cost recovery are suggested.

  15. Virtual environments for nuclear power plant design

    SciTech Connect (OSTI)

    Brown-VanHoozer, S.A.; Singleterry, R.C. Jr.; King, R.W. [and others

    1996-03-01T23:59:59.000Z

    In the design and operation of nuclear power plants, the visualization process inherent in virtual environments (VE) allows for abstract design concepts to be made concrete and simulated without using a physical mock-up. This helps reduce the time and effort required to design and understand the system, thus providing the design team with a less complicated arrangement. Also, the outcome of human interactions with the components and system can be minimized through various testing of scenarios in real-time without the threat of injury to the user or damage to the equipment. If implemented, this will lead to a minimal total design and construction effort for nuclear power plants (NPP).

  16. Floating nuclear power plant safety assurance principles

    SciTech Connect (OSTI)

    Zvonarev, B.M.; Kuchin, N.L.; Sergeev, I.V.

    1993-12-31T23:59:59.000Z

    In the north regions of the Russian federation and low density population areas, there is a real necessity for ecological clean energy small power sources. For this purpose, floating nuclear power plants, designed on the basis of atomic ship building engineering, are being conceptualized. It is possible to use the ship building plants for the reactor purposes. Issues such as radioactive waste management are described.

  17. Microsoft Word - PC-000580_0_NGNP Technology Road Mapping Report...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    lower reactor outlet helium temperature range. NGNP Technology Development Road Mapping Report PC-0005800 ix TABLE OF CONTENTS ACRONYMS ......

  18. Microsoft Word - 911168_0 NGNP Fuel Specification Basis_rel.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    ... 19 2.3 Interrelationship with NGNP Core Performance Analysis Task ... 20 2.4 Background and Planned Approach...

  19. SELFMONITORING DISTRIBUTED MONITORING SYSTEM FOR NUCLEAR POWER PLANTS (PRELIMINARY VERSION)

    E-Print Network [OSTI]

    SELF­MONITORING DISTRIBUTED MONITORING SYSTEM FOR NUCLEAR POWER PLANTS (PRELIMINARY VERSION) Aldo and identification are extremely important activities for the safety of a nuclear power plant. In particular inside huge and complex production plants. 1 INTRODUCTION Safety in nuclear power plants requires

  20. Nuclear plant irradiated steel handbook

    SciTech Connect (OSTI)

    Oldfield, W.; Oldfield, F.M.; Lombrozo, P.M.; McConnell, P.

    1986-09-01T23:59:59.000Z

    This reference handbook presents selected information extracted from the EPRI reactor surveillance program database, which contains the results from surveillance program reports on 57 plants and 116 capsules. Tabulated data includes radiation induced temperature shifts, capsule irradiation conditions and statistical features of the Charpy V-notch curves. General information on the surveillance materials is provided and the Charpy V-notch energy results are presented graphically.

  1. Date Set for Closure of Russian Nuclear Weapons Plant - NNSA...

    National Nuclear Security Administration (NNSA)

    Date Set for Closure of Russian Nuclear Weapons Plant - NNSA Is Helping Make It Happen | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission...

  2. Microsoft Word - 911167_0 Pt Design Rpt_rel.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    7 Revision 0 ENGINEERING SERVICES FOR THE NEXT GENERATION NUCLEAR PLANT (NGNP) WITH HYDROGEN PRODUCTION Nuclear Heat Supply System Point Design Study for NGNP Conceptual Design...

  3. SIGNAL GROUPING FOR CONDITION MONITORING OF NUCLEAR POWER PLANT COMPONENTS

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    SIGNAL GROUPING FOR CONDITION MONITORING OF NUCLEAR POWER PLANT COMPONENTS Piero Baraldi Chevalier EDF R&D ­ Simulation and information Technologies for Power generation system Department 6, Quai Monitoring, Empirical Modeling, Power Plants, Safety Critical Nuclear Instrumentation, Autoassociative models

  4. Construction or Extended Operation of Nuclear Plant (Vermont)

    Broader source: Energy.gov [DOE]

    Any petition for approval of construction of a nuclear energy generating plant within the state, or any petition for approval of the operation of a nuclear energy generating plant beyond the date...

  5. Design and Status of the NGNP Fuel Experiment AGR-3/4 Irradiated in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Blaine Grover

    2012-10-01T23:59:59.000Z

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in November 2013. Since the purpose of this experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is significantly different from the first two experiments, though the control and monitoring systems are very similar. The purpose and design of this experiment will be discussed followed by its progress and status to date.

  6. Microsoft Word - Final Report on NGNP Construction at Contaminated...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    unit at the Grand Gulf site in Mississippi. Entergy prepared and submitted to the NRC two combined license applications for new nuclear plants at two different sites - through...

  7. Microsoft Word - 11-21605_Nuclear_Facility_Boundary_White_Paper...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    those specified for the nuclear - industrial facility boundary. g. Functions from NGNP white paper, INLEXT-10-19509, "Structures, Systems, and Components Safety Classification,"...

  8. Advanced nuclear plant control room complex

    DOE Patents [OSTI]

    Scarola, Kenneth (Windsor, CT); Jamison, David S. (Windsor, CT); Manazir, Richard M. (North Canton, CT); Rescorl, Robert L. (Vernon, CT); Harmon, Daryl L. (Enfield, CT)

    1993-01-01T23:59:59.000Z

    An advanced control room complex for a nuclear power plant, including a discrete indicator and alarm system (72) which is nuclear qualified for rapid response to changes in plant parameters and a component control system (64) which together provide a discrete monitoring and control capability at a panel (14-22, 26, 28) in the control room (10). A separate data processing system (70), which need not be nuclear qualified, provides integrated and overview information to the control room and to each panel, through CRTs (84) and a large, overhead integrated process status overview board (24). The discrete indicator and alarm system (72) and the data processing system (70) receive inputs from common plant sensors and validate the sensor outputs to arrive at a representative value of the parameter for use by the operator during both normal and accident conditions, thereby avoiding the need for him to assimilate data from each sensor individually. The integrated process status board (24) is at the apex of an information hierarchy that extends through four levels and provides access at each panel to the full display hierarchy. The control room panels are preferably of a modular construction, permitting the definition of inputs and outputs, the man machine interface, and the plant specific algorithms, to proceed in parallel with the fabrication of the panels, the installation of the equipment and the generic testing thereof.

  9. Improved Economics of Nuclear Plant Life Management

    SciTech Connect (OSTI)

    Bond, Leonard J.; Doctor, Steven R.; Jarrell, Donald B.; Bond, Joseph W D.

    2007-07-31T23:59:59.000Z

    The adoption of new on-line monitoring, diagnostic and eventually prognostics technologies has the potential to impact the economics of the existing nuclear power plant fleet, new plants and future advanced designs. To move from periodic inspection to on-line monitoring for condition based maintenance and eventually prognostics will require advances in sensors, better understanding of what and how to measure within the plant; enhanced data interrogation, communication and integration; new predictive models for damage/aging evolution; system integration for real world deployments; quantification of uncertainties in what are inherently ill-posed problems and integration of enhanced condition based maintenance/prognostics philosophies into new plant designs, operation and O&M approaches. The move to digital systems in petrochemical, process and fossil fuel power plants is enabling major advances to occur in the instrumentation, controls and monitoring systems and approaches employed. The adoption within the nuclear power community of advanced on-line monitoring and advanced diagnostics has the potential for the reduction in costly periodic surveillance that requires plant shut-down , more accurate cost-benefit analysis, “just-in-time” maintenance, pre-staging of maintenance tasks, move towards true “operation without failures” and a jump start on advanced technologies for new plant concepts, such as those under the International Gen IV Program. There are significant opportunities to adopt condition-based maintenance when upgrades are implemented at existing facilities. The economic benefit from a predictive maintenance program based upon advanced on-line monitoring and advanced diagnostics can be demonstrated from a cost/benefit analysis. An analysis of the 104 US legacy systems has indicated potential savings at over $1B per year when applied to all key equipment; a summary of the supporting analysis is provided in this paper.

  10. Seismic requirements for design of nuclear power plants and nuclear test facilities

    SciTech Connect (OSTI)

    Not Available

    1985-02-01T23:59:59.000Z

    This standard establishes engineering requirements for the design of nuclear power plants and nuclear test facilities to accommodate vibratory effects of earthquakes.

  11. Nuclear power plant performance assessment pertaining to plant aging in France and the United States

    E-Print Network [OSTI]

    Guyer, Brittany (Brittany Leigh)

    2013-01-01T23:59:59.000Z

    The effect of aging on nuclear power plant performance has come under increased scrutiny in recent years. The approaches used to make an assessment of this effect strongly influence the economics of nuclear power plant ...

  12. Cesium Removal at Fukushima Nuclear Plant - 13215

    SciTech Connect (OSTI)

    Braun, James L.; Barker, Tracy A. [Avantech Incorporated, 95A Sunbelt Blvd Columbia, SC 29203 (United States)] [Avantech Incorporated, 95A Sunbelt Blvd Columbia, SC 29203 (United States)

    2013-07-01T23:59:59.000Z

    The Great East Japan Earthquake that took place on March 11, 2011 created a number of technical challenges at the Fukushima Daiichi Nuclear Plant. One of the primary challenges involved the treatment of highly contaminated radioactive wastewater. Avantech Inc. developed a unique patent pending treatment system that addressed the numerous technical issues in an efficient and safe manner. Our paper will address the development of the process from concept through detailed design, identify the lessons learned, and provide the updated results of the project. Specific design and operational parameters/benefits discussed in the paper include: - Selection of equipment to address radionuclide issues; - Unique method of solving the additional technical issues associated with Hydrogen Generation and Residual Heat; - Operational results, including chemistry, offsite discharges and waste generation. Results show that the customized process has enabled the utility to recycle the wastewater for cooling and reuse. This technology had a direct benefit to nuclear facilities worldwide. (authors)

  13. Plutonium Processing Plant Deactivated | National Nuclear Security

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 - September 2006Photovoltaic TheoryPlant 242-Z Americium Recovery Facility

  14. Microsoft Word - 911106_0 End-ProductsReport_final_rel.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    6 Revision 0 Preconceptual Engineering Services For The Next Generation Nuclear Plant (NGNP) With Hydrogen Production NGNP End-Products Study Prepared by General Atomics For the...

  15. 911104_0 final_rel.pdf

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    04 Revision 0 Preconceptual Engineering Services For The Next Generation Nuclear Plant (NGNP) With Hydrogen Production NGNP Reactor Power Level Study Prepared by General Atomics...

  16. Microsoft Word - 911125_0_rel.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    5 Revision 0 ENGINEERING SERVICES FOR THE NEXT GENERATION NUCLEAR PLANT (NGNP) WITH HYDROGEN PRODUCTION NGNP Composites R&D Technical Issues Study Prepared by General Atomics For...

  17. Microsoft Word - Appd K GA_Exec_Summ_Flysheet.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    44 Revision 0 Preconceptual Engineering Services For The Next Generation Nuclear Plant (NGNP) With Hydrogen Production Executive Summary Report - NGNP and Hydrogen Production...

  18. Microsoft Word - 911102_0 SRM_rel.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    2 Revision 0 Preconceptual Engineering Services For The Next Generation Nuclear Plant (NGNP) With Hydrogen Production NGNP System Requirements Manual Prepared by General Atomics...

  19. Microsoft Word - 911119_a IHX & HTS Alternatives-Judy1-04-23...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    9 Revision 0 Engineering Services for the Next Generation Nuclear Plant (NGNP) with Hydrogen Production NGNP IHX and Secondary Heat Transport Loop Alternatives Study Prepared by...

  20. 911105_0 final_rel.pdf

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    5 Revision 0 Preconceptual Engineering Services For The Next Generation Nuclear Plant (NGNP) With Hydrogen Production NGNP High Temperature Process Heat Transfer and Transport...

  1. Microsoft Word - 911120_0_SG Rpt_rel.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Revision 0 Engineering Services for the Next Generation Nuclear Plant (NGNP) with Hydrogen Production NGNP Steam Generator Alternatives Study Prepared by General Atomics For the...

  2. Microsoft Word - 911160_0 CPA Phase 1 report _FINAL__rel.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    0 Revision 0 ENGINEERING SERVICES FOR THE NEXT GENERATION NUCLEAR PLANT (NGNP) WITH HYDROGEN PRODUCTION Final Report - NGNP Core Performance Analysis, Phase 1 Prepared by General...

  3. Microsoft Word - 911117_a contamination control-Judy1-04-23-08...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    17 Revision 0 Engineering Services For The Next Generation Nuclear Plant (NGNP) With Hydrogen Production NGNP Contamination Control Study Prepared by General Atomics For the...

  4. Microsoft Word - 911127_0 Parametric-Calculations_rel.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    7 Revision 0 Engineering Services for the Next Generation Nuclear Plant (NGNP) with Hydrogen Production NGNP Parametric Fuel and Reactor Pressure Vessel Temperature Calculations...

  5. Microsoft Word - PC-000566_0_White Paper_rel.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Services for the Next Generation Nuclear Plant (NGNP) with Hydrogen Production White Paper - Characterizing the Effect of NGNP Operating Conditions on the Uncertainty of...

  6. News Release Closure of Russian Nuclear Plant.PDF

    National Nuclear Security Administration (NNSA)

    RELEASE Jonathan Kiell, 202586-7371 September 27, 2001 Date Set for Closure of Russian Nuclear Weapons Plant U.S. National Nuclear Security Administration Is Helping Make It...

  7. Microsoft Word - 12-9075581-000 NGNP Risk Paper r05.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    NGNP Risk Evaluation of Major Components Document No. 12-9075581-000 Page 12 such as gas compressors. Large motors certainly exist. Variable frequency power supplies exist. The...

  8. Electromagnetic Compatibility in Nuclear Power Plants

    SciTech Connect (OSTI)

    Ewing, P.D.; Kercel, S.W.; Korsah, K.; Wood, R.T.

    1999-08-29T23:59:59.000Z

    Electromagnetic compatibility (EMC) has long been a key element of qualification for mission critical instrumentation and control (I&C) systems used by the U.S. military. The potential for disruption of safety-related I&C systems by electromagnetic interference (EMI), radio-frequency interference (RFI), or power surges is also an issue of concern for the nuclear industry. Experimental investigations of the potential vulnerability of advanced safety systems to EMI/RFI, coupled with studies of reported events at nuclear power plants (NPPs) that are attributed to EMI/RFI, confirm the safety significance of EMC for both analog and digital technology. As a result, Oak Ridge National Laboratory has been engaged in the development of the technical basis for guidance that addresses EMC for safety-related I&C systems in NPPs. This research has involved the identification of engineering practices to minimize the potential impact of EMI/RFI and power surges and an evaluation of the ambient electromagnetic environment at NPPs to tailor those practices for use by the nuclear industry. Recommendations for EMC guidance have been derived from these research findings and are summarized in this paper.

  9. Risk-informed incident management for nuclear power plants

    E-Print Network [OSTI]

    Smith, Curtis Lee, 1966-

    2002-01-01T23:59:59.000Z

    Decision making as a part of nuclear power plant operations is a critical, but common, task. Plant management is forced to make decisions that may have safety and economic consequences. Formal decision theory offers the ...

  10. Sensitivity analysis for the outages of nuclear power plants

    E-Print Network [OSTI]

    Kengy Barty

    2012-02-17T23:59:59.000Z

    Feb 17, 2012 ... Abstract: Nuclear power plants must be regularly shut down in order to perform refueling and maintenance operations. The scheduling of the ...

  11. Optimization Online - Nuclear norm minimization for the planted ...

    E-Print Network [OSTI]

    Brendan Ames

    2009-01-21T23:59:59.000Z

    Jan 21, 2009 ... Nuclear norm minimization for the planted clique and biclique problems. Brendan Ames(bpames ***at*** math.uwaterloo.ca) Stephen ...

  12. RADIOLOGICAL HEALTH AND RELATED STANDARDS FOR NUCLEAR POWER PLANTS. VOLUME 2 OF HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01T23:59:59.000Z

    average value for nuclear plants) aFinal Envir. Statement (Statement, Koshkonong Nuclear Plant, August 1976. U. S.rem; operation of the nuclear plants themselves only *Other

  13. Scaling studies and conceptual experiment designs for NGNP CFD assessment

    SciTech Connect (OSTI)

    D. M. McEligot; G. E. McCreery

    2004-11-01T23:59:59.000Z

    The objective of this report is to document scaling studies and conceptual designs for flow and heat transfer experiments intended to assess CFD codes and their turbulence models proposed for application to prismatic NGNP concepts. The general approach of the project is to develop new benchmark experiments for assessment in parallel with CFD and coupled CFD/systems code calculations for the same geometry. Two aspects of the complex flow in an NGNP are being addressed: (1) flow and thermal mixing in the lower plenum ("hot streaking" issue) and (2) turbulence and resulting temperature distributions in reactor cooling channels ("hot channel" issue). Current prismatic NGNP concepts are being examined to identify their proposed flow conditions and geometries over the range from normal operation to decay heat removal in a pressurized cooldown. Approximate analyses have been applied to determine key non-dimensional parameters and their magnitudes over this operating range. For normal operation, the flow in the coolant channels can be considered to be dominant turbulent forced convection with slight transverse property variation. In a pressurized cooldown (LOFA) simulation, the flow quickly becomes laminar with some possible buoyancy influences. The flow in the lower plenum can locally be considered to be a situation of multiple hot jets into a confined crossflow -- with obstructions. Flow is expected to be turbulent with momentumdominated turbulent jets entering; buoyancy influences are estimated to be negligible in normal full power operation. Experiments are needed for the combined features of the lower plenum flows. Missing from the typical jet experiments available are interactions with nearby circular posts and with vertical posts in the vicinity of vertical walls - with near stagnant surroundings at one extreme and significant crossflow at the other. Two types of heat transfer experiments are being considered. One addresses the "hot channel" problem, if necessary. The second type will treat heated jets entering a model plenum. Unheated MIR (Matched-Index-of-Refraction) experiments are first steps when the geometry is complicated. One does not want to use a computational technique which will not even handle constant properties properly. The purpose of the fluid dynamics experiments is to develop benchmark databases for the assessment of CFD solutions of the momentum equations, scalar mixing and turbulence models for typical NGNP plenum geometries in the limiting case of negligible buoyancy and constant fluid properties. As indicated by the scaling studies, in normal full power operation of a typical NGNP conceptual design, buoyancy influences should be negligible in the lower plenum. The MIR experiment will simulate flow features of the paths of jets as they mix in flowing through the array of posts in a lower plenum en route to the single exit duct. Conceptual designs for such experiments are described.

  14. Inspection of Nuclear Power Plant Containment Structures

    SciTech Connect (OSTI)

    Graves, H.L.; Naus, D.J.; Norris, W.E.

    1998-12-01T23:59:59.000Z

    Safety-related nuclear power plant (NPP) structures are designed to withstand loadings from a number of low-probability external and interval events, such as earthquakes, tornadoes, and loss-of-coolant accidents. Loadings incurred during normal plant operation therefore generally are not significant enough to cause appreciable degradation. However, these structures are susceptible to aging by various processes depending on the operating environment and service conditions. The effects of these processes may accumulate within these structures over time to cause failure under design conditions, or lead to costly repair. In the late 1980s and early 1990s several occurrences of degradation of NPP structures were discovered at various facilities (e.g., corrosion of pressure boundary components, freeze- thaw damage of concrete, and larger than anticipated loss of prestressing force). Despite these degradation occurrences and a trend for an increasing rate of occurrence, in-service inspection of the safety-related structures continued to be performed in a somewhat cursory manner. Starting in 1991, the U.S. Nuclear Regulatory Commission (USNRC) published the first of several new requirements to help ensure that adequate in-service inspection of these structures is performed. Current regulatory in-service inspection requirements are reviewed and a summary of degradation experience presented. Nondestructive examination techniques commonly used to inspect the NPP steel and concrete structures to identify and quantify the amount of damage present are reviewed. Finally, areas where nondestructive evaluation techniques require development (i.e., inaccessible portions of the containment pressure boundary, and thick heavily reinforced concrete sections are discussed.

  15. RADIOLOGICAL EMERGENCY RESPONSE PLANNING FOR NUCLEAR POWER PLANTS IN CALIFORNIA. VOLUME 4 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Yen, W.W.S.

    2010-01-01T23:59:59.000Z

    EMERGENCY PLANNING FOR NUCLEAR POWER PLANTS: THE LICENSINGEmergency Planning for Nuclear Power Plants Determination ofproposed nuclear power plants . . . . . . . . . • . . . .

  16. Some aspects of the decommissioning of nuclear power plants

    SciTech Connect (OSTI)

    Khvostova, M. S., E-mail: marinakhvostova@list.ru [St. Petersburg State Maritime Technical University (Sevmashvtuz), Severodvinsk Branch (Russian Federation)

    2012-03-15T23:59:59.000Z

    The major factors influencing the choice of a national concept for the decommissioning of nuclear power plants are examined. The operating lifetimes of power generating units with nuclear reactors of various types (VVER-1000, VVER-440, RBMK-1000, EGP-6, and BN-600) are analyzed. The basic approaches to decommissioning Russian nuclear power plants and the treatment of radioactive waste and spent nuclear fuel are discussed. Major aspects of the ecological and radiation safety of personnel, surrounding populations, and the environment during decommissioning of nuclear installations are identified.

  17. The Daya Bay Nuclear Plant Project in the Light of International Environmental Law

    E-Print Network [OSTI]

    Mushkat, Roda

    1990-01-01T23:59:59.000Z

    result from locating a nuclear plant so close to the Hongat 1292 (1975). THE DA YA BAY NUCLEAR PLANT PROJECT national1986) (H.K. ). THE DA YA BAY NUCLEAR PLANT PROJECT IV. THE "

  18. Design of the Next Generation Nuclear Plant Graphite Creep Experiments for Irradiation in the Advanced Test Reactor

    SciTech Connect (OSTI)

    S. Blaine Grover

    2009-05-01T23:59:59.000Z

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain seven separate stacks of graphite specimens. Six of the specimen stacks will have half of their graphite specimens under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will be organized into pairs with a different compressive load being applied to the top half of each pair of specimen stacks. The seventh stack will not have a compressive load on the graphite specimens during irradiation. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during initial start-up of the experiment. The final design phase for the first experiment was completed in September 2008, and the fabrication and assembly of the experiment test train as well as installation and testing of the control and support systems that will monitor and control the experiment during irradiation are being completed in early calendar 2009. The first experiment is scheduled to be ready for insertion in the ATR by April 30, 2009. This paper will discuss the design of the experiment including the test train and the temperature and compressive load monitoring, control, and data collection systems.

  19. Risk Framework for the Next Generation Nuclear Power Plant Construction

    E-Print Network [OSTI]

    Yeon, Jaeheum 1981-

    2012-12-11T23:59:59.000Z

    sector projects, and recently elevated to Best Practice status. However, its current format is inadequate to address the unique challenges of constructing the next generation of nuclear power plants (NPP). To understand and determine the risks...

  20. Mapping complexity sources in nuclear power plant domains

    E-Print Network [OSTI]

    Sasangohar, Farzan

    Understanding the sources of complexity in advanced Nuclear Power Plant (NPP) control rooms and their effects on human reliability is critical for ensuring safe performance of both operators and the entire system. New ...

  1. Can New Nuclear Power Plants be Project Financed?

    E-Print Network [OSTI]

    Taylor, Simon

    This paper considers the prospects for financing a wave of new nuclear power plants (NPP) using project financing, which is used widely in large capital intensive infrastructure investments, including the power and gas sectors, but has...

  2. The Daya Bay Nuclear Plant Project in the Light of International Environmental Law

    E-Print Network [OSTI]

    Mushkat, Roda

    1990-01-01T23:59:59.000Z

    Ministry of Nuclear Industry; PACIFIC BASIN LAW JOURNAL [international law prohibits a state from building a nuclearNUCLEAR PLANT PROJECT IN THE LIGHT OF INTERNATIONAL ENVIRONMENTAL LAW

  3. Diffusion Welding of Compact Heat Exchangers for Nuclear Applications

    SciTech Connect (OSTI)

    Denis Clark; Ron Mizia; Dr. Michael V. Glazoff; Mr. Michael W. Patterson

    2012-06-01T23:59:59.000Z

    The next-­-generation nuclear plant (NGNP) is designed to be a flexible source of energy, producing various mixes of electrical energy and process heat (for example, for hydrogen generation) on demand. Compact heat exchangers provide an attractive way to move energy from the helium primary reactor coolant to process heat uses. For process heat efficiency, reactor outlet temperatures of 750-­-900°C are desirable. There are minor but deleterious components in the primary coolant; the number of alloys that can handle this environment is small. The present work concentrates on Alloys 800H and 617.

  4. Development of decontamination techniques for decommissioning commercial nuclear power plants

    SciTech Connect (OSTI)

    Ishikura, T.; Miwa, T.; Onozawa, T.; Ohtsuka, H. [Nuclear Power Engineering Corp., Tokyo (Japan). Plant and Components Dept.; Ishigure, K. [Univ. of Tokyo (Japan). Dept. of Quantum Engineering and System Science

    1993-12-31T23:59:59.000Z

    NUPEC has been developing various techniques to safely and efficiently decommission large commercial nuclear power plants. The development work, referred to as the verification tests, has been performed since 1982. The verification tests on decontamination techniques have focused on the reduction of both occupational radiation exposure and radioactive waste volume. Experiments on various decontamination methods have been carried out. Prospects of applying efficient decontamination techniques to commercial nuclear power plant decommissioning are bright due to the experimental results.

  5. UNSUPERVISED CLUSTERING FOR FAULT DIAGNOSIS IN NUCLEAR POWER PLANT COMPONENTS

    E-Print Network [OSTI]

    Boyer, Edmond

    1 UNSUPERVISED CLUSTERING FOR FAULT DIAGNOSIS IN NUCLEAR POWER PLANT COMPONENTS Piero Baraldi1 of prototypical behaviors. Its performance is tested with respect to an artificial case study and then applied on transients originated by different faults in the pressurizer of a nuclear power reactor. Key Words: Fault

  6. Boron control system for a nuclear power plant

    SciTech Connect (OSTI)

    Brown, W.W.; Van der Schoot, M.R.

    1980-09-30T23:59:59.000Z

    Ion exchangers which reversibly store borate ions in a temperature dependent process are combined with evaporative boric acid recovery apparatus to provide a boron control system for controlling the reactivity of nuclear power plants. A plurality of ion exchangers are operated sequentially to provide varying amounts of boric acid to a nuclear reactor for load follow operations. Evaporative boric acid recovery apparatus is utilized for major changes in the boron concentration within the nuclear reactor.

  7. Next Generation Nuclear Plant Licensing Strategy

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    reactor that is based on research and development (R&D) activities supported by the Generation IV Nuclear Energy Systems Initiative and shall be used to generate electricity,...

  8. ISOLATION OF NUCLEAR DNA FROM PLANTS Based on Peterson et al. (1997), Plant Mol. Biol. Reptr. 15: 148-153.

    E-Print Network [OSTI]

    Ray, David

    1997-01-01T23:59:59.000Z

    ISOLATION OF NUCLEAR DNA FROM PLANTS Based on Peterson et al. (1997), Plant Mol. Biol. Reptr. 15 quantities of nuclear DNA from a wide variety of plants including pine, tomato, juniper, cypress, sorghum for plants in which polyphenols are a problem, although it has provided good results for every plant species

  9. Aging of concrete structures in nuclear power plants

    SciTech Connect (OSTI)

    Naus, D.J.; Pland, C.B. (Oak Ridge National Lab., TN (USA)); Arndt, E.G. (Nuclear Regulatory Commission, Washington, DC (USA))

    1991-01-01T23:59:59.000Z

    The Structural Aging (SAG) Program, sponsored by the US Nuclear Regulatory Commission (USNRC) and conducted by the Oak Ridge National Laboratory (ORNL), had the overall objective of providing the USNRC with an improved basis for evaluating nuclear power plant structures for continued service. The program consists of three technical tasks: materials property data base, structural component assessment/repair technology, and quantitative methodology for continued service determinations. Major accomplishments under the SAG Program during the first two years of its planned five-year duration have included: development of a Structural Materials Information Center and formulation of a Structural Aging Assessment Methodology for Concrete Structures in Nuclear Power Plants. 9 refs.

  10. Aging management of containment structures in nuclear power plants

    SciTech Connect (OSTI)

    Naus, D.J.; Oland, C.B. [Oak Ridge National Lab., TN (United States); Ellingwood, B.R. [The Johns Hopkins Univ., Baltimore, MD (United States); Graves, H.L. III; Norris, W.E. [US Nuclear Regulatory Commission, Washington, DC (United States)

    1994-12-31T23:59:59.000Z

    Research is being conducted by ORNL under US Nuclear Regulatory Commission (USNRC) sponsorship to address aging management of nuclear power plant containment and other safety-related structures. Documentation is being prepared to provide the USNRC with potential structural safety issues and acceptance criteria for use in continued service evaluations of nuclear power plants. Accomplishments include development of a Structural Materials Information Center containing data and information on the time variation of 144 material properties under the influence of pertinent environmental stressors or aging factors, evaluation of models for potential concrete containment degradation factors, development of a procedure to identify critical structures and degradation factors important to aging management, evaluations of nondestructive evaluation techniques. assessments of European and North American repair practices for concrete, review of parameters affecting corrosion of metals embedded in concrete, and development of methodologies for making current condition assessments and service life predictions of new or existing reinforced concrete structures in nuclear power plants.

  11. Tour of Entergy's Nuclear Power Plant in River Bend Owner: Entergy Gulf States Inc.

    E-Print Network [OSTI]

    Ervin, Elizabeth K.

    Tour of Entergy's Nuclear Power Plant in River Bend Owner: Entergy Gulf States Inc. Reactor Type a nuclear power plant. Plant was Entergy, a Boiling Water Reactor (BWR) type. Built in the 80's, it has of the veteran plant workers. The presentation gave the nuclear plant engineering basics and built

  12. Microsoft Word - NGNP PCDR_Rev1_16Nov07.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    technology reference as the base. The shortcomings of achieving this mission with a small-scale NGNP test reactor (approximately 25-50 MWt) are the following. The proposed...

  13. Microsoft Word - INL NGNP-HTGR Final Report with INL Comments...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    o Coal to Methanol to Gasoline x TEV 671 o Coal to Substitute Natural Gas x TEV 672 o Coal to Liquids o Gas to Liquids x TEV 674 Date: 2212011 NGNP-HTGR Assisted Conventional...

  14. Regulatory guidance for lightning protection in nuclear power plants

    SciTech Connect (OSTI)

    Kisner, R. A.; Wilgen, J. B.; Ewing, P. D.; Korsah, K. [Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, TN 37831-6007 (United States); Antonescu, C. E. [U.S. Nuclear Regulatory Commission, Washington, DC 20555 (United States)

    2006-07-01T23:59:59.000Z

    Oak Ridge National Laboratory (ORNL) was engaged by the U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear Regulatory Research (RES) to develop the technical basis for regulatory guidance to address design and implementation practices for lightning protection systems in nuclear power plants (NPPs). Lightning protection is becoming increasingly important with the advent of digital and low-voltage analog systems in NPPs. These systems have the potential to be more vulnerable than older analog systems to the resulting power surges and electromagnetic interference (EMI) when lightning strikes facilities or power lines. This paper discusses the technical basis for guidance to licensees and applicants covered in Regulatory Guide (RG) 1.204, Guidelines for Lightning Protection of Nuclear Power Plants, issued August 2005. RG 1.204 describes guidance for practices that are acceptable to the NRC staff for protecting nuclear power structures and systems from direct lightning strikes and the resulting secondary effects. (authors)

  15. Regulatory Guidance for Lightning Protection in Nuclear Power Plants

    SciTech Connect (OSTI)

    Kisner, Roger A [ORNL; Wilgen, John B [ORNL; Ewing, Paul D [ORNL; Korsah, Kofi [ORNL; Antonescu, Christina E [ORNL

    2006-01-01T23:59:59.000Z

    Abstract - Oak Ridge National Laboratory (ORNL) was engaged by the U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear Regulatory Research (RES) to develop the technical basis for regulatory guidance to address design and implementation practices for lightning protection systems in nuclear power plants (NPPs). Lightning protection is becoming increasingly important with the advent of digital and low-voltage analog systems in NPPs. These systems have the potential to be more vulnerable than older analog systems to the resulting power surges and electromagnetic interference (EMI) when lightning strikes facilities or power lines. This paper discusses the technical basis for guidance to licensees and applicants covered in Regulatory Guide (RG) 1.204, Guidelines for Lightning Protection of Nuclear Power Plants, issued August 2005. RG 1.204 describes guidance for practices that are acceptable to the NRC staff for protecting nuclear power structures and systems from direct lightning strikes and the resulting secondary effects.

  16. New Hampshire Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    total reactors","Summer capacity (nw)","Net generation (thousand mwh)","Share of State nuclear net generation (percent)","Owner" "Seabrook Unit 1","1,247","10,910",100.0,"NextEr...

  17. Summary of NGNP Engineering Analysis Progress in Fiscal Year 2010

    SciTech Connect (OSTI)

    Phillip M. Mills

    2010-09-01T23:59:59.000Z

    The Engineering Analysis work package accomplished a number of activities in FY2010 that will help to inform Licensing, R&D, and detailed design activities that will be performed during the preliminary design phase in Phase 2 of the NGNP Project. These activities were in the following areas: • Fission Product Transport • DDN Update and Consolidation • Gas Reactor Lessons Learned Review • Reactor Coolant Chemistry Control • Resilient Control Systems for High Temperature Gas-cooled Reactors • Water-ingress Analysis In total, the efforts associated with the Engineering Analysis work package accomplished four (4) Level 2 milestones and two (2) internal (Level 4) milestones. Details of the activities and milestones are included in the attached report.

  18. US nuclear power plant operating cost and experience summaries

    SciTech Connect (OSTI)

    Kohn, W.E.; Reid, R.L.; White, V.S.

    1998-02-01T23:59:59.000Z

    NUREG/CR-6577, U.S. Nuclear Power Plant Operating Cost and Experience Summaries, has been prepared to provide historical operating cost and experience information on U.S. commercial nuclear power plants. Cost incurred after initial construction are characterized as annual production costs, representing fuel and plant operating and maintenance expenses, and capital expenditures related to facility additions/modifications which are included in the plant capital asset base. As discussed in the report, annual data for these two cost categories were obtained from publicly available reports and must be accepted as having different degrees of accuracy and completeness. Treatment of inconclusive and incomplete data is discussed. As an aid to understanding the fluctuations in the cost histories, operating summaries for each nuclear unit are provided. The intent of these summaries is to identify important operating events; refueling, major maintenance, and other significant outages; operating milestones; and significant licensing or enforcement actions. Information used in the summaries is condensed from annual operating reports submitted by the licensees, plant histories contained in Nuclear Power Experience, trade press articles, and the Nuclear Regulatory Commission (NRC) web site (www.nrc.gov).

  19. Nuclear Power Plant NDE Challenges - Past, Present, and Future

    SciTech Connect (OSTI)

    Doctor, S. R. [Pacific Northwest National Laboratory, P.O. Box 999, Richland, WA 99352 (United States)

    2007-03-21T23:59:59.000Z

    The operating fleet of U.S. nuclear power plants was built to fossil plant standards (of workmanship, not fitness for service) and with good engineering judgment. Fortuitously, those nuclear power plants were designed using defense-in-depth concepts, with nondestructive examination (NDE) an important layer, so they can tolerate almost any component failure and still continue to operate safely. In the 30+ years of reactor operation, many material failures have occurred. Unfortunately, NDE has not provided the reliability to detect degradation prior to initial failure (breaching the pressure boundary). However, NDE programs have been improved by moving from prescriptive procedures to performance demonstrations that quantify inspection effectiveness for flaw detection probability and sizing accuracy. Other improvements include the use of risk-informed strategies to ensure that reactor components contributing the most risk receive the best and most frequent inspections. Another challenge is the recent surge of interest in building new nuclear power plants in the United States to meet increasing domestic energy demand. New construction will increase the demand for NDE but also offers the opportunity for more proactive inspections. This paper reviews the inception and evolution of NDE for nuclear power plants over the past 40 years, recounts lessons learned, and describes the needs remaining as existing plants continue operation and new construction is contemplated.

  20. IEEE TRANSACTIONS ON NUCLEAR SCIENCE, VOL. 58, NO. 1, FEBRUARY 2011 277 Anomaly Detection in Nuclear Power Plants via

    E-Print Network [OSTI]

    Ray, Asok

    in Nuclear Power Plants via Symbolic Dynamic Filtering Xin Jin, Student Member, IEEE, Yin Guo, Soumik Sarkar detection algorithm for condition monitoring of nuclear power plants, where symbolic feature extraction Innova- tive & Secure (IRIS) simulator of nuclear power plants, and its per- formance is evaluated

  1. Nuclear Power Plant Containment Pressure Boundary Research

    SciTech Connect (OSTI)

    Cherry, J.L.; Chokshi, N.C.; Costello, J.F.; Ellingwood, B.R.; Naus, D.J.

    1999-09-15T23:59:59.000Z

    Research to address aging of the containment pressure boundary in light-water reactor plants is summarized. This research is aimed at understanding the significant factors relating occurrence of corrosion, efficacy of inspection, and structural capacity reduction of steel containment and liners of concrete containment. This understanding will lead to improvements in risk-informed regulatory decision making. Containment pressure boundary components are described and potential aging factors identified. Quantitative tools for condition assessments of aging structures to maintain an acceptable level of reliability over the service life of the plant are discussed. Finally, the impact of aging (i.e., loss of shell thickness due to corrosion) on steel containment fragility for a pressurized water reactor ice-condenser plant is presented.

  2. Status of Initial Assessment of Physical and Mechanical Properties of Graphite Grades for NGNP Appkications

    SciTech Connect (OSTI)

    Strizak, Joe P [ORNL; Burchell, Timothy D [ORNL; Windes, Will [Idaho National Laboratory (INL)

    2011-12-01T23:59:59.000Z

    Current candidate graphite grades for the core structures of NGNP include grades NBG-17, NBG-18, PCEA and IG-430. Both NBG-17 and NBG-18 are manufactured using pitch coke, and are vibrationally molded. These medium grain products are produced by SGL Carbon SAS (France). Tayo Tanso (Japan) produces IG-430 which is a petroleum coke, isostatically molded, nuclear grade graphite. And PCEA is a medium grain, extruded graphite produced by UCAR Carbon Co. (USA) from petroleum coke. An experimental program has been initiated to develop physical and mechanical properties data for these current candidate graphites. The results will be judged against the requirements for nuclear grade graphites set forth in ASTM standard D 7219-05 "Standard Specification for Isotropic and Near-isotropic Nuclear Graphites". Physical properties data including thermal conductivity and coefficient of thermal expansion, and mechanical properties data including tensile, compressive and flexural strengths will be obtained using the established test methods covered in D-7219 and ASTM C 781-02 "Standard Practice for Testing Graphite and Boronated Graphite Components for High-Temperature Gas-Cooled Nuclear Reactors". Various factors known to effect the properties of graphites will be investigated. These include specimen size, spatial location within a graphite billet, specimen orientation (ag and wg) within a billet, and billet-to-billet variations. The current status of the materials characterization program is reported herein. To date billets of the four graphite grades have been procured, and detailed cut up plans for obtaining the various specimens have been prepared. Particular attention has been given to the traceability of each specimen to its spatial location and orientation within a billet.

  3. Service experience and reliability improvement: Nuclear, fossil, and petrochemical plants

    SciTech Connect (OSTI)

    Bamford, W.H.; Cipolla, R.C.; Warke, W.R.; Onyewuenyi, O.A.; Bagnoli, D.; Phillips, J.H.; Prager, M.; Becht, C. IV (eds.)

    1994-01-01T23:59:59.000Z

    This publication contains papers presented at the following four symposia conducted at the 1994 Pressure Vessels and Piping Conference in Minneapolis, Minnesota, June 19--23: Service Experience in Nuclear Plants; Risk-Based Inspection and Evaluation; Service Experience in Operating Fossil Power Plants; and Service Experience in Petrochemical Plants. These symposia were sponsored by the Materials and Fabrication and the Design and Analysis Committees of the ASME Pressure Vessels and Piping Division. The objective of these symposia was to disseminate information on issues and degradation that have resulted from the operation of nuclear, fossil, and petrochemical power plants, as well as related reliability issues. Thirty-nine papers have been processed separately for inclusion on the data base.

  4. Service experience and life management: Nuclear, fossil, and petrochemical plants

    SciTech Connect (OSTI)

    Bamford, W.H. (ed.)

    1993-01-01T23:59:59.000Z

    This publication contains papers presented at four symposia conducted at the 1993 Pressure Vessels and Piping Conference in Denver, Colorado, July 25--29. The symposia titles are listed below: Service Experience and Reliability Improvement in Nuclear Plants; Service Experience in Operating Fossil Power Plants; Service Experience in Petrochemical Plants; Aging Management, Condition Monitoring and Diagnostics. These symposia were sponsored by the Materials and Fabrication and the Design and Analysis Committees of the ASME Pressure Vessels and Piping Division. The objective of these sessions was to disseminate information on issues and degradation which have resulted from the operation of nuclear, fossil, and petrochemical power plants, as well as related monitoring and diagnostic techniques. Individual papers have been processed separately for inclusion in the appropriate data bases.

  5. Report on aging of nuclear power plant reinforced concrete structures

    SciTech Connect (OSTI)

    Naus, D.J.; Oland, C.B. [Oak Ridge National Lab., TN (United States); Ellingwood, B.R. [Johns Hopkins Univ., Baltimore, MD (United States). Dept. of Civil Engineering

    1996-03-01T23:59:59.000Z

    The Structural Aging Program provides the US Nuclear Regulatory Commission with potential structural safety issues and acceptance criteria for use in continued service assessments of nuclear power plant safety-related concrete structures. The program was organized under four task areas: Program Management, Materials Property Data Base, Structural Component Assessment/Repair Technology, and Quantitative Methodology for Continued Service Determinations. Under these tasks, over 90 papers and reports were prepared addressing pertinent aspects associated with aging management of nuclear power plant reinforced concrete structures. Contained in this report is a summary of program results in the form of information related to longevity of nuclear power plant reinforced concrete structures, a Structural Materials Information Center presenting data and information on the time variation of concrete materials under the influence of environmental stressors and aging factors, in-service inspection and condition assessments techniques, repair materials and methods, evaluation of nuclear power plant reinforced concrete structures, and a reliability-based methodology for current and future condition assessments. Recommendations for future activities are also provided. 308 refs., 61 figs., 50 tabs.

  6. Nuclear power plant fire protection: philosophy and analysis. [PWR; BWR

    SciTech Connect (OSTI)

    Berry, D. L.

    1980-05-01T23:59:59.000Z

    This report combines a fire severity analysis technique with a fault tree methodology for assessing the importance to nuclear power plant safety of certain combinations of components and systems. Characteristics unique to fire, such as propagation induced by the failure of barriers, have been incorporated into the methodology. By applying the resulting fire analysis technique to actual conditions found in a representative nuclear power plant, it is found that some safety and nonsafety areas are both highly vulnerable to fire spread and impotant to overall safety, while other areas prove to be of marginal importance. Suggestions are made for further experimental and analytical work to supplement the fire analysis method.

  7. A methodology for evaluating ``new`` technologies in nuclear power plants

    SciTech Connect (OSTI)

    Korsah, K.; Clark, R.L.; Holcomb, D.E.

    1994-06-01T23:59:59.000Z

    As obsolescence and spare parts issues drive nuclear power plants to upgrade with new technology (such as optical fiber communication systems), the ability of the new technology to withstand stressors present where it is installed needs to be determined. In particular, new standards may be required to address qualification criteria and their application to the nuclear power plants of tomorrow. This paper discusses the failure modes and age-related degradation mechanisms of fiber optic communication systems, and suggests a methodology for identifying when accelerated aging should be performed during qualification testing.

  8. Reproductive Life Events in the Population Living in the Vicinity of a Nuclear Waste Reprocessing Plant

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    : There is concern about the health of populations living close to nuclear waste reprocessing plants. We conducted the health of the population living in the vicinity of nuclear waste reprocessing plants was raised the Dounreay nuclear waste reprocessing plant (United Kingdom).[1] Similar studies around the French nuclear

  9. Yeast-Plant Coupled Vector System for Identification of Nuclear Proteins1[OA

    E-Print Network [OSTI]

    Citovsky, Vitaly

    Yeast-Plant Coupled Vector System for Identification of Nuclear Proteins1[OA] Adi Zaltsman, Bu.G.) Nuclear proteins are involved in many critical biological processes within plant cells and, therefore nuclear localization. Thus, studies of plant nuclear proteins would be facilitated by a convenient

  10. Ground-based testing of space nuclear power plants

    SciTech Connect (OSTI)

    McDonald, T.G.

    1990-10-22T23:59:59.000Z

    Small nuclear power plants for space applications are evaluated according to their testability in this two part report. The first part introduces the issues involved in testing these power plants. Some of the concerns include oxygen embrittlement of critical components, the test environment, the effects of a vacuum environment on materials, the practically of racing an activated test chamber, and possible testing alternative the SEHPTR, king develop at the Idaho National Engineering Laboratory. 10 refs., 6 figs., 1 tab.

  11. The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

    SciTech Connect (OSTI)

    S. Blaine Grover

    2009-09-01T23:59:59.000Z

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In addition, the purpose and differences between the two experiments will be compared and the irradiation results to date on the first experiment will be presented.

  12. ASSESSING NUCLEAR POWER PLANT SAFETY AND RECOVERY FROM EARTHQUAKES USING A SYSTEM-OF-SYSTEMS

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    by Monte Carlo simulation the probability that the nuclear power plant enters in an unsafe stateASSESSING NUCLEAR POWER PLANT SAFETY AND RECOVERY FROM EARTHQUAKES USING A SYSTEM in which the plant is embedded. As a test system, we consider the impacts produced on a nuclear power plant

  13. A formal software requirements specification method for digital nuclear plant protection systems

    E-Print Network [OSTI]

    A formal software requirements specification method for digital nuclear plant protection systems plant protection system in nuclear power plants. NuSCR improves the readability and specifiability those in aerospace, satellite and nuclear power plants, whose failure could result in danger to human

  14. Aging assessment of large electric motors in nuclear power plants

    SciTech Connect (OSTI)

    Villaran, M.; Subudhi, M. [Brookhaven National Lab., Upton, NY (United States)

    1996-03-01T23:59:59.000Z

    Large electric motors serve as the prime movers to drive high capacity pumps, fans, compressors, and generators in a variety of nuclear plant systems. This study examined the stressors that cause degradation and aging in large electric motors operating in various plant locations and environments. The operating history of these machines in nuclear plant service was studied by review and analysis of failure reports in the NPRDS and LER databases. This was supplemented by a review of motor designs, and their nuclear and balance of plant applications, in order to characterize the failure mechanisms that cause degradation, aging, and failure in large electric motors. A generic failure modes and effects analysis for large squirrel cage induction motors was performed to identify the degradation and aging mechanisms affecting various components of these large motors, the failure modes that result, and their effects upon the function of the motor. The effects of large motor failures upon the systems in which they are operating, and on the plant as a whole, were analyzed from failure reports in the databases. The effectiveness of the industry`s large motor maintenance programs was assessed based upon the failure reports in the databases and reviews of plant maintenance procedures and programs.

  15. Liquid metal cooled nuclear reactor plant system

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

    1993-01-01T23:59:59.000Z

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

  16. RADIOLOGICAL HEALTH AND RELATED STANDARDS FOR NUCLEAR POWER PLANTS. VOLUME 2 OF HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01T23:59:59.000Z

    in U. S. Conunercial Nuclear Power Plants", Report WASH-Related Standards for Nuclear Power Plants," by A.V. NeroResponse Planning for Nuclear Power Plants in California,"

  17. Naval Nuclear Propulsion Plants | National Nuclear Security Administration

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsrucLas Conchas recoveryLaboratory | NationalJohn Cyber Security NuclearNewNatural

  18. Naval Nuclear Propulsion Plants | National Nuclear Security Administration

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsrucLas Conchas recoveryLaboratory | NationalJohn Cyber Security NuclearNewNaturalOur Mission /

  19. Aging management guideline for commercial nuclear power plants - heat exchangers

    SciTech Connect (OSTI)

    Booker, S.; Lehnert, D.; Daavettila, N.; Palop, E.

    1994-06-01T23:59:59.000Z

    This Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in commercial nuclear power plant heat exchangers important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.

  20. Kansas City Plant | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartmentNational Nuclear SecurityNationalApply for Our Jobs /Operations / Acquisition and

  1. Radioactive Effluents from Nuclear Power Plants Annual Report 2007

    SciTech Connect (OSTI)

    U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation

    2010-12-10T23:59:59.000Z

    This report describes radioactive effluents from commercial nuclear power plants (NPPs) in the United States. This information was reported by the licensees for radioactive discharges that occurred in 2007. The report provides information relevant to the potential impact of NPPs on the environment and on public health.

  2. Radioactive Effluents from Nuclear Power Plants Annual Report 2008

    SciTech Connect (OSTI)

    U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation

    2010-12-10T23:59:59.000Z

    This report describes radioactive effluents from commercial nuclear power plants (NPPs) in the United States. This information was reported by the licensees for radioactive discharges that occurred in 2008. The report provides information relevant to the potential impact of NPPs on the environment and on public health.

  3. RADIOLOGICAL EMERGENCY RESPONSE PLANNING FOR NUCLEAR POWER PLANTS IN CALIFORNIA. VOLUME 4 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Yen, W.W.S.

    2010-01-01T23:59:59.000Z

    S. Commercial Nuclear Power Plants. WASH-1400. October 1975.Content of for Nuclear Power Plants. Regulatory Guide 1.101.PLANNING FOR NUCLEAR POWER PLANTS: THE LICENSING PROCESS

  4. RADIOLOGICAL EMERGENCY RESPONSE PLANNING FOR NUCLEAR POWER PLANTS IN CALIFORNIA. VOLUME 4 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Yen, W.W.S.

    2010-01-01T23:59:59.000Z

    from the Rancho Seco nuclear plant was simulated, A total ofdistributions around the nuclear plant sites based on thegrowth surrounding nuclear plants after the issuance of the

  5. CONTROL OF POPULATION DENSITIES SURROUNDING NUCLEAR POWER PLANTS. VOLUME 5 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, jA.V.

    2010-01-01T23:59:59.000Z

    surrounding a nuclear plant, and they are stronglylocation for a nuclear plant, but it is the measures thatand consequences of nuclear plant accidents and would match

  6. Enhancement of NRC station blackout requirements for nuclear power plants

    SciTech Connect (OSTI)

    McConnell, M. W. [United States Nuclear Regulatory Commission, Mail Stop: 012-H2, Washington, DC 20555 (United States)

    2012-07-01T23:59:59.000Z

    The U.S. Nuclear Regulatory Commission (NRC) established a Near-Term Task Force (NTTF) in response to Commission direction to conduct a systematic and methodical review of NRC processes and regulations to determine whether the agency should make additional improvements to its regulatory system and to make recommendations to the Commission for its policy direction, in light of the accident at the Fukushima Dai-ichi Nuclear Power Plant. The NTTF's review resulted in a set of recommendations that took a balanced approach to defense-in-depth as applied to low-likelihood, high-consequence events such as prolonged station blackout (SBO) resulting from severe natural phenomena. Part 50, Section 63, of Title 10 of the Code of Federal Regulations (CFR), 'Loss of All Alternating Current Power,' currently requires that each nuclear power plant must be able to cool the reactor core and maintain containment integrity for a specified duration of an SBO. The SBO duration and mitigation strategy for each nuclear power plant is site specific and is based on the robustness of the local transmission system and the transmission system operator's capability to restore offsite power to the nuclear power plant. With regard to SBO, the NTTF recommended that the NRC strengthen SBO mitigation capability at all operating and new reactors for design-basis and beyond-design-basis external events. The NTTF also recommended strengthening emergency preparedness for prolonged SBO and multi-unit events. These recommendations, taken together, are intended to clarify and strengthen US nuclear reactor safety regarding protection against and mitigation of the consequences of natural disasters and emergency preparedness during SBO. The focus of this paper is on the existing SBO requirements and NRC initiatives to strengthen SBO capability at all operating and new reactors to address prolonged SBO stemming from design-basis and beyond-design-basis external events. The NRC initiatives are intended to enhance core and spent fuel pool cooling, reactor coolant system integrity, and containment integrity. (authors)

  7. Online Monitoring of Plant Assets in the Nuclear Industry

    SciTech Connect (OSTI)

    Nancy Lybeck; Vivek Agarwal; Binh Pham; Richard Rusaw; Randy Bickford

    2013-10-01T23:59:59.000Z

    Today’s online monitoring technologies provide opportunities to perform predictive and proactive health management of assets within many different industries, in particular the defense and aerospace industries. The nuclear industry can leverage these technologies to enhance safety, productivity, and reliability of the aging fleet of existing nuclear power plants. The U.S. Department of Energy’s Light Water Reactor Sustainability Program is collaborating with the Electric Power Research Institute’s (EPRI’s) Long-Term Operations program to implement online monitoring in existing nuclear power plants. Proactive online monitoring in the nuclear industry is being explored using EPRI’s Fleet-Wide Prognostic and Health Management (FW-PHM) Suite software, a set of web-based diagnostic and prognostic tools and databases that serves as an integrated health monitoring architecture. This paper focuses on development of asset fault signatures used to assess the health status of generator step-up transformers and emergency diesel generators in nuclear power plants. Asset fault signatures describe the distinctive features based on technical examinations that can be used to detect a specific fault type. Fault signatures are developed based on the results of detailed technical research and on the knowledge and experience of technical experts. The Diagnostic Advisor of the FW-PHM Suite software matches developed fault signatures with operational data to provide early identification of critical faults and troubleshooting advice that could be used to distinguish between faults with similar symptoms. This research is important as it will support the automation of predictive online monitoring techniques in nuclear power plants to diagnose incipient faults, perform proactive maintenance, and estimate the remaining useful life of assets.

  8. Reprocessing of nuclear fuels at the Savannah River Plant

    SciTech Connect (OSTI)

    Gray, L.W.

    1986-10-04T23:59:59.000Z

    For more than 30 years, the Savannah River Plant (SRP) has been a major supplier of nuclear materials such as plutonium-239 and tritium-3 for nuclear and thermonuclear weapons, plutonium-238 for space exploration, and isotopes of americium, curium, and californium for use in the nuclear research community. SRP is a complete nuclear park, providing most of the processes in the nuclear fuel cycle. Key processes involve fabrication and cladding of the nuclear fuel, target, and control assemblies; rework of heavy water for use as reactor moderator; reactor loading, operation, and unloading; chemical recovery of the reactor transmutation products and spent fuels; and management of the gaseous, liquid, and solid nuclear and chemical wastes; plus a host of support operations. The site's history and the key processes from fabrication of reactor fuels and targets to finishing of virgin plutonium for use in the nuclear weapons complex are reviewed. Emphasis has been given to the chemistry of the recovery and purification of weapons grade plutonium from irradiated reactor targets.

  9. Operating nuclear plant feedback to ASME and French codes

    SciTech Connect (OSTI)

    Journet, J. [Electricite de France, Clamart (France); O`Donnell, W.J. [O`Donnell Consulting Engineers, Bethel Park, PA (United States)

    1996-12-01T23:59:59.000Z

    The French have an advantage in nuclear plant operating experience feedback due to the highly centralized nature of their nuclear industry. There is only one utility in charge of design as well as operations (EDF) and only one reactor vendor (Framatome). The ASME Code has played a key role in resolving technical issues in the design and operation of nuclear plants since the inception of nuclear power. The committee structure of the Code brings an ideal combination of senior technical people with both broad and specialized experience to bear on complex how safe is safe enough technical issues. The authors now see an even greater role for the ASME Code in a proposed new regulatory era for the US nuclear industry. The current legalistic confrontational regulatory era has been quite destructive. There now appears to be a real opportunity to begin a new era of technical consensus as the primary means for resolving safety issues. This change can quickly be brought about by having the industry take operating plant problems and regulatory technical issues directly to the ASME Code for timely resolution. Surprisingly, there is no institution in the US nuclear industry with such a mandate. In fact, the industry is organized to feedback through the Nuclear Regulatory Commission issues which could be far better resolved through the ASME Code. Major regulatory benefits can be achieved by closing this loop and providing systematic interaction with the ASME Code. The essential elements of a new regulatory era and ideas for organizing US institutional industry responsibilities, taken from the French experience, are described in this paper.

  10. Mapping Complexity Sources in Nuclear Power Plant Domains Understanding the sources of complexity in advanced Nuclear Power Plant (NPP) control rooms and their

    E-Print Network [OSTI]

    Cummings, Mary "Missy"

    Mapping Complexity Sources in Nuclear Power Plant Domains Understanding the sources of complexity in advanced Nuclear Power Plant (NPP) control rooms and their effects on human reliability is critical of complexity leveraging network theory. INTRODUCTION The nuclear power industry in United States has declined

  11. A Roadmap to Deploy New Nuclear Power Plants in the United States...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    A Roadmap to Deploy New Nuclear Power Plants in the United States by 2010: Volume II, Main Report A Roadmap to Deploy New Nuclear Power Plants in the United States by 2010: Volume...

  12. Innovative applications of technology for nuclear power plant productivity improvements

    SciTech Connect (OSTI)

    Naser, J. A. [Electric Power Research Inst., 3420 Hillview Avenue, Palo Alto, CA 94303 (United States)

    2012-07-01T23:59:59.000Z

    The nuclear power industry in several countries is concerned about the ability to maintain high plant performance levels due to aging and obsolescence, knowledge drain, fewer plant staff, and new requirements and commitments. Current plant operations are labor-intensive due to the vast number of operational and support activities required by commonly used technology in most plants. These concerns increase as plants extend their operating life. In addition, there is the goal to further improve performance while reducing human errors and increasingly focus on reducing operations and maintenance costs. New plants are expected to perform more productively than current plants. In order to achieve and increase high productivity, it is necessary to look at innovative applications of modern technologies and new concepts of operation. The Electric Power Research Inst. is exploring and demonstrating modern technologies that enable cost-effectively maintaining current performance levels and shifts to even higher performance levels, as well as provide tools for high performance in new plants. Several modern technologies being explored can provide multiple benefits for a wide range of applications. Examples of these technologies include simulation, visualization, automation, human cognitive engineering, and information and communications technologies. Some applications using modern technologies are described. (authors)

  13. Requirements for advanced simulation of nuclear reactor and chemicalseparation plants.

    SciTech Connect (OSTI)

    Palmiotti, G.; Cahalan, J.; Pfeiffer, P.; Sofu, T.; Taiwo, T.; Wei,T.; Yacout, A.; Yang, W.; Siegel, A.; Insepov, Z.; Anitescu, M.; Hovland,P.; Pereira, C.; Regalbuto, M.; Copple, J.; Willamson, M.

    2006-12-11T23:59:59.000Z

    This report presents requirements for advanced simulation of nuclear reactor and chemical processing plants that are of interest to the Global Nuclear Energy Partnership (GNEP) initiative. Justification for advanced simulation and some examples of grand challenges that will benefit from it are provided. An integrated software tool that has its main components, whenever possible based on first principles, is proposed as possible future approach for dealing with the complex problems linked to the simulation of nuclear reactor and chemical processing plants. The main benefits that are associated with a better integrated simulation have been identified as: a reduction of design margins, a decrease of the number of experiments in support of the design process, a shortening of the developmental design cycle, and a better understanding of the physical phenomena and the related underlying fundamental processes. For each component of the proposed integrated software tool, background information, functional requirements, current tools and approach, and proposed future approaches have been provided. Whenever possible, current uncertainties have been quoted and existing limitations have been presented. Desired target accuracies with associated benefits to the different aspects of the nuclear reactor and chemical processing plants were also given. In many cases the possible gains associated with a better simulation have been identified, quantified, and translated into economical benefits.

  14. Indicator system for advanced nuclear plant control complex

    DOE Patents [OSTI]

    Scarola, Kenneth (Windsor, CT); Jamison, David S. (Windsor, CT); Manazir, Richard M. (North Canton, CT); Rescorl, Robert L. (Vernon, CT); Harmon, Daryl L. (Enfield, CT)

    1993-01-01T23:59:59.000Z

    An advanced control room complex for a nuclear power plant, including a discrete indicator and alarm system (72) which is nuclear qualified for rapid response to changes in plant parameters and a component control system (64) which together provide a discrete monitoring and control capability at a panel (14-22, 26, 28) in the control room (10). A separate data processing system (70), which need not be nuclear qualified, provides integrated and overview information to the control room and to each panel, through CRTs (84) and a large, overhead integrated process status overview board (24). The discrete indicator and alarm system (72) and the data processing system (70) receive inputs from common plant sensors and validate the sensor outputs to arrive at a representative value of the parameter for use by the operator during both normal and accident conditions, thereby avoiding the need for him to assimilate data from each sensor individually. The integrated process status board (24) is at the apex of an information hierarchy that extends through four levels and provides access at each panel to the full display hierarchy. The control room panels are preferably of a modular construction, permitting the definition of inputs and outputs, the man machine interface, and the plant specific algorithms, to proceed in parallel with the fabrication of the panels, the installation of the equipment and the generic testing thereof.

  15. Targeting of PlantTargeting of Plant RanGAPRanGAP to the Nuclear Envelopeto the Nuclear Envelope Annkatrin Rose,Annkatrin Rose, ShalakaShalaka S. Patel, Iris MeierS. Patel, Iris Meier

    E-Print Network [OSTI]

    Meier, Iris

    Targeting of PlantTargeting of Plant RanGAPRanGAP to the Nuclear Envelopeto the Nuclear Envelope RanGAP1 and tomato MAF1. Plant RanGAP and MAF1 are targeted to the nuclear envelope in plant cells to be cytoplasmic. Plant RanGAP contains a N- terminal domain shared with the nuclear envelope protein MAF1 (cyan

  16. Requirements for Advanced Simulation of Nuclear Reactor and Chemical Separation Plants

    E-Print Network [OSTI]

    Anitescu, Mihai

    Requirements for Advanced Simulation of Nuclear Reactor and Chemical Separation Plants ANL-AFCI-168 of Nuclear Reactor and Chemical Separation Plants ANL-AFCI-168 by G. Palmiotti, J. Cahalan, P. Pfeiffer, T;2 ANL-AFCI-168 Requirements for Advanced Simulation of Nuclear Reactor and Chemical Separation Plants G

  17. A Domain-Specific Safety Analysis for Digital Nuclear Plant Protection Systems

    E-Print Network [OSTI]

    A Domain-Specific Safety Analysis for Digital Nuclear Plant Protection Systems Sanghyun Yoon through safety analy- sis is strongly mandated for safety-critical systems. Nuclear plant protection, NuFTA, for nuclear plant protection systems. NuFTA mechanically constructs a software fault tree

  18. A Verification Framework for FBD based Software in Nuclear Power Plants Junbeom Yoo

    E-Print Network [OSTI]

    A Verification Framework for FBD based Software in Nuclear Power Plants Junbeom Yoo Div, conducted using a nuclear power plant shutdown system being developed in Korea, demonstrated in nuclear power plant's reactor protection systems. The software verification framework uses two different

  19. Formal Verification of Safety I&C System Designs: Two Nuclear Power Plant Related Applications

    E-Print Network [OSTI]

    Heljanko, Keijo

    C4.2 Formal Verification of Safety I&C System Designs: Two Nuclear Power Plant Related Applications and control (I&C) systems play a crucial role in the operation of nuclear power plants (NPP) and other safety is available. The use of model checking to verify two nuclear power plant related systems is described: an arc

  20. Nuclear Power Plant Components Condition Monitoring by Probabilistic Support Vector , Redouane Seraouib

    E-Print Network [OSTI]

    Boyer, Edmond

    Nuclear Power Plant Components Condition Monitoring by Probabilistic Support Vector Machine Jie.zio@ecp.fr Abstract In this paper, an approach for the prediction of the condition of Nuclear Power Plant (NPP monitoring, Nuclear power plant, Point prediction hal-00790421,version1-12Jun2013 Author manuscript

  1. ATP-dependent regulation of nuclear Ca2 levels in plant cells

    E-Print Network [OSTI]

    Shaw, Peter

    ATP-dependent regulation of nuclear Ca2 levels in plant cells Tom D. Bunney, Peter J. Shaw, Peter A in [Ca2+ ] occurs in the nuclear periphery. The occurrence of ATP-dependent Ca2+ uptake in plant nuclei rights reserved. Key words: Nucleus; Plant; Ca2 uptake; Signal transduction; Imaging; Nuclear pore

  2. Evolution of a Visual Impact Model to Evaluate Nuclear Plant Siting and Design Option1

    E-Print Network [OSTI]

    Standiford, Richard B.

    Evolution of a Visual Impact Model to Evaluate Nuclear Plant Siting and Design Option1 2/ Brian A and economic options for the analysis of nuclear plant siting possibilities (Burnham 1974; Jones, April 1975 of nuclear plant siting options for the AEC. BNWL's multi-disciplinary impact evaluation pro- cedure required

  3. PLC-Based Safety Critical Software Development for Nuclear Power Plants

    E-Print Network [OSTI]

    PLC-Based Safety Critical Software Development for Nuclear Power Plants Junbeom Yoo1 , Sungdeok Cha development technique for nuclear power plants'I&C soft- ware controllers. To improve software safety, we in developing safety-critical control software for a Korean nuclear power plant, and experience to date has been

  4. Condition monitoring of motor-operated valves in nuclear power plants Pierre Granjon

    E-Print Network [OSTI]

    Boyer, Edmond

    Condition monitoring of motor-operated valves in nuclear power plants Pierre Granjon Gipsa of nuclear power plants. Unfortunately, today's policies present a major drawback. Indeed, these monitoring safety constraints: nuclear power plants. Key components of such systems are motor-operated valves (MOVs

  5. Vulnerability Analysis of a Nuclear Power Plant Considering Detonations of Explosive Devices

    E-Print Network [OSTI]

    Cizelj, Leon

    Vulnerability Analysis of a Nuclear Power Plant Considering Detonations of Explosive Devices Marko threats to a nuclear power plant in the year 1991 and after the 9/11 events in 2001. The methodology which strength and injuries of human beings with nuclear power plant models used in probabilistic safety

  6. Childhood leukaemia incidence below the age of 5 years near French nuclear power plants

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Childhood leukaemia incidence below the age of 5 years near French nuclear power plants D Laurier 1 living in the vicinity of nuclear power plants in Germany. We present herein results about the incidence of childhood leukaemia in the vicinity of nuclear power plants in France for the same age range. These results

  7. U.S. Nuclear Power Plants: Continued Life or Replacement After 60? (released in AEO2010)

    Reports and Publications (EIA)

    2010-01-01T23:59:59.000Z

    Nuclear power plants generate approximately 20% of U.S. electricity, and the plants in operation today are often seen as attractive assets in the current environment of uncertainty about future fossil fuel prices, high construction costs for new power plants (particularly nuclear plants), and the potential enactment of greenhouse gas regulations. Existing nuclear power plants have low fuel costs and relatively high power output. However, there is uncertainty about how long they will be allowed to continue operating.

  8. Review of maintenance personnel practices at nuclear power plants

    SciTech Connect (OSTI)

    Chockie, A.D.; Badalamente, R.V.; Hostick, C.J.; Vickroy, S.C.; Bryant, J.L.; Imhoff, C.H.

    1984-05-01T23:59:59.000Z

    As part of the Nuclear Regulatory Commission (NRC) sponsored Maintenance Qualifications and Staffing Project, the Pacific Northwest Laboratory (PNL) has conducted a preliminary assessment of nuclear power plant (NPP) maintenance practices. As requested by the NRC, the following areas within the maintenance function were examined: personnel qualifications, maintenance training, overtime, shiftwork and staffing levels. The purpose of the assessment was to identify the primary safety-related problems that required further analysis before specific recommendations can be made on the regulations affecting NPP maintenance operations.

  9. Infrastructure development assistance modeling for nuclear power plant

    SciTech Connect (OSTI)

    Park, J. H.; Hwang, K.; Park, K. M.; Kim, S. W.; Lee, S. M. [Korea Hydro and Nuclear Power Co., LTD, 23, 106 gil, Yeongdong-daero, Gangnam-gu, 153-791 (Korea, Republic of)

    2012-07-01T23:59:59.000Z

    The purpose of this paper is to develop a model, a general frame to be utilized in assisting newcomer countries to start a nuclear power program. A nuclear power plant project involves technical complexity and high level of investment with long duration. Considering newcomers are mostly developing countries that lack the national infrastructure, key infrastructure issues may constitute the principal constraints to the development of a nuclear power program. In this regard, it is important to provide guidance and support to set up an appropriate infrastructure when we help them with the first launch of nuclear power plant project. To date, as a sole nuclear power generation company, KHNP has been invited many times to mentor or assist newcomer countries for their successful start of a nuclear power program since Republic of Korea is an exemplary case of a developing country which began nuclear power program from scratch and became a major world nuclear energy country in a short period of time. Through hosting events organized to aid newcomer countries' initiation of nuclear power projects, difficulties have been recognized. Each event had different contents according to circumstances because they were held as an unstructured and one-off thing. By developing a general model, we can give more adequate and effective aid in an efficient way. In this paper, we created a model to identify necessary infrastructures at the right stage, which was mainly based on a case of Korea. Taking into account the assistance we received from foreign companies and our own efforts for technological self-reliance, we have developed a general time table and specified activities required to do at each stage. From a donor's perspective, we explored various ways to help nuclear infrastructure development including technical support programs, training courses, and participating in IAEA technical cooperation programs on a regular basis. If we further develop the model, the next task would be to make the model more sophisticated as a 'semi-tailored model' so that it can be applied to a certain country reflecting its unique conditions. In accordance with its degree of established infrastructure, we can adjust or modify the model. Despite lots of benefits of using this model, there remain limitations such as time and budget constraints. These problems, however, can be addressed by cooperating with international organization such as the IAEA and other companies that share the same goal of helping newcomer countries introduce nuclear power. (authors)

  10. Summary for the Next Generation Nuclear Plant Project in Review

    SciTech Connect (OSTI)

    L.E. Demick

    2010-09-01T23:59:59.000Z

    This paper reports on the major progress that the NGNP Project has made toward developing and commercializing the HTGR technology. Significant R&D progress has been made in addressing key technical issues for qualification of the HTGR fuel and graphite, codification of high temperature materials and verification and validation of design codes. Work is also progressing in heat transfer/transport design and testing and in development of the high temperature steam electrolysis hydrogen production process. A viable licensing strategy has been formulated in coordination with the NRC and DOE. White papers covering key licensing issues have been and will continue to be submitted and necessary discussions of these key issues have begun with the NRC. Continued government support is needed to complete the Project objectives as established in the 2005 Energy Policy Act.

  11. Decommissioning nuclear power plants - the wave of the future

    SciTech Connect (OSTI)

    Griggs, F.S. Jr. [Raytheon Engineers and Contractors, Cumberland City, TN (United States)

    1994-12-31T23:59:59.000Z

    The paper discusses the project controls developed in the decommissioning of a nuclear power plant. Considerations are given to the contaminated piping and equipment that have to be removed and the spent and used fuel that has to be disposed of. The storage issue is of primary concern here. The cost control aspects and the dynamics of decommissioning are discussed. The effects of decommissioning laws on the construction and engineering firms are mentioned. 5 refs.

  12. Human Factors Considerations in New Nuclear Power Plants: Detailed Analysis.

    SciTech Connect (OSTI)

    OHara,J.; Higgins, J.; Brown, W.; Fink, R.

    2008-02-14T23:59:59.000Z

    This Nuclear Regulatory Commission (NRC) sponsored study has identified human-performance issues in new and advanced nuclear power plants. To identify the issues, current industry developments and trends were evaluated in the areas of reactor technology, instrumentation and control technology, human-system integration technology, and human factors engineering (HFE) methods and tools. The issues were organized into seven high-level HFE topic areas: Role of Personnel and Automation, Staffing and Training, Normal Operations Management, Disturbance and Emergency Management, Maintenance and Change Management, Plant Design and Construction, and HFE Methods and Tools. The issues where then prioritized into four categories using a 'Phenomena Identification and Ranking Table' methodology based on evaluations provided by 14 independent subject matter experts. The subject matter experts were knowledgeable in a variety of disciplines. Vendors, utilities, research organizations and regulators all participated. Twenty issues were categorized into the top priority category. This Brookhaven National Laboratory (BNL) technical report provides the detailed methodology, issue analysis, and results. A summary of the results of this study can be found in NUREG/CR-6947. The research performed for this project has identified a large number of human-performance issues for new control stations and new nuclear power plant designs. The information gathered in this project can serve as input to the development of a long-term strategy and plan for addressing human performance in these areas through regulatory research. Addressing human-performance issues will provide the technical basis from which regulatory review guidance can be developed to meet these challenges. The availability of this review guidance will help set clear expectations for how the NRC staff will evaluate new designs, reduce regulatory uncertainty, and provide a well-defined path to new nuclear power plant licensing.

  13. Understanding the nature of nuclear power plant risk

    SciTech Connect (OSTI)

    Denning, R. S. [Ohio State Univ., 201 West 19th Avenue, Columbus, OH 43210-1142 (United States)

    2012-07-01T23:59:59.000Z

    This paper describes the evolution of understanding of severe accident consequences from the non-mechanistic assumptions of WASH-740 to WASH-1400, NUREG-1150, SOARCA and today in the interpretation of the consequences of the accident at Fukushima. As opposed to the general perception, the radiological human health consequences to members of the Japanese public from the Fukushima accident will be small despite meltdowns at three reactors and loss of containment integrity. In contrast, the radiation-related societal impacts present a substantial additional economic burden on top of the monumental task of economic recovery from the nonnuclear aspects of the earthquake and tsunami damage. The Fukushima accident provides additional evidence that we have mis-characterized the risk of nuclear power plant accidents to ourselves and to the public. The human health risks are extremely small even to people living next door to a nuclear power plant. The principal risk associated with a nuclear power plant accident involves societal impacts: relocation of people, loss of land use, loss of contaminated products, decontamination costs and the need for replacement power. Although two of the three probabilistic safety goals of the NRC address societal risk, the associated quantitative health objectives in reality only address individual human health risk. This paper describes the types of analysis that would address compliance with the societal goals. (authors)

  14. Conceivable new recycling of nuclear waste by nuclear power companies in their plants

    E-Print Network [OSTI]

    Ruggero Maria Santilli

    1997-04-09T23:59:59.000Z

    We outline the basic principles and the needed experiments for a conceivable new recycling of nuclear waste by the power plants themselves to avoid its transportation and storage to a (yet unknown) dumping area. Details are provided in an adjoining paper and in patents pending.

  15. NGNP Data Management and Analysis System Analysis and Web Delivery Capabilities

    SciTech Connect (OSTI)

    Cynthia D. Gentillon

    2011-09-01T23:59:59.000Z

    Projects for the Very High Temperature Reactor (VHTR) 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. The NGNP Data Management and Analysis System (NDMAS) at the Idaho National Laboratory has been established to ensure that VHTR 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. The capabilities are described from the perspective of NDMAS users, starting with those who just view experimental data and analytical results on the INL NDMAS web portal. Web display and delivery capabilities are described in detail. Also the current web pages that show Advanced Gas Reactor, Advanced Graphite Capsule, and High Temperature Materials test results are itemized. Capabilities available to NDMAS developers are more extensive, and are described using a second series of examples. Much of the data analysis efforts focus on understanding how thermocouple measurements relate to simulated temperatures and other experimental parameters. Statistical control charts and correlation monitoring provide an ongoing assessment of instrument accuracy. Data analysis capabilities are virtually unlimited for those who use the NDMAS web data download capabilities and the analysis software of their choice. Overall, the NDMAS provides convenient data analysis and web delivery capabilities for studying a very large and rapidly increasing database of well-documented, pedigreed data.

  16. The Meteorological Monitoring program at a former nuclear weapons plant

    SciTech Connect (OSTI)

    Maxwell, D.R.; Bowen, B.M.

    1994-02-01T23:59:59.000Z

    The purpose of the Meteorological Monitoring program at Rocky Flats Plant (RFP) is to provide meteorological information for use in assessing the transport, and diffusion, and deposition of effluent actually or potentially released into the atmosphere by plant operations. Achievement of this objective aids in protecting health and safety of the public, employees, and environment, and directly supports Emergency Response programs at RFP. Meteorological information supports the design of environmental monitoring networks for impact assessments, environmental surveillance activities, remediation activities, and emergency responses. As the mission of the plant changes from production of nuclear weapons parts to environmental cleanup and economic development, smaller releases resulting from remediation activities become more likely. These possible releases could result from airborne fugitive dust, evaporation from collection ponds, or grass fires.

  17. Aging management guideline for commercial nuclear power plants-pumps

    SciTech Connect (OSTI)

    Booker, S.; Katz, D.; Daavettila, N.; Lehnert, D. [MDC-Ogden Environmental and Energy Services, Southfield, MI (United States)

    1994-03-01T23:59:59.000Z

    This Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant pumps important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.

  18. The Plant Cell, Vol. 10, 16371647, October 1998, www.plantcell.org 1998 American Society of Plant Physiologists The Plant U1 Small Nuclear Ribonucleoprotein Particle

    E-Print Network [OSTI]

    Reddy, A.S.N

    of Plant Physiologists The Plant U1 Small Nuclear Ribonucleoprotein Particle 70K Protein Interacts with TwoThe Plant Cell, Vol. 10, 1637­1647, October 1998, www.plantcell.org © 1998 American Society small nuclear ribonucleoprotein particle (U1 snRNP) 70K protein (U1-70K), one of the three U1 sn

  19. RADIOLOGICAL HEALTH AND RELATED STANDARDS FOR NUCLEAR POWER PLANTS. VOLUME 2 OF HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01T23:59:59.000Z

    Report LBL-5287. "Power Plant Reliability-Availability andConunercial Nuclear Power Plants", Report WASH-1400 (NUREG-Standards for Nuclear Power Plants," by A.V. Nero and Y.C.

  20. Compiling Utility Requirements For New Nuclear Power Plant Project

    SciTech Connect (OSTI)

    Patrakka, Eero [Teollisuuden Voima Oy, 27160 Olkiluoto (Finland)

    2002-07-01T23:59:59.000Z

    Teollisuuden Voima Oy (TVO) submitted in November 2000 to the Finnish Government an application for a Decision-in-Principle concerning the construction of a new nuclear power plant in Finland. The actual investment decision can be made first after a positive decision has been made by the Government and the Parliament. Parallel to the licensing process, technical preparedness has been upheld so that the procurement process can be commenced without delay, when needed. This includes the definition of requirements for the plant and preliminary preparation of bid inquiry specifications. The core of the technical requirements corresponds to the specifications presented in the European Utility Requirement (EUR) document, compiled by major European electricity producers. Quite naturally, an amount of modifications to the EUR document are needed that take into account the country- and site-specific conditions as well as the experiences gained in the operation of the existing NPP units. Along with the EUR-related requirements concerning the nuclear island and power generation plant, requirements are specified for scope of supply as well as for a variety of issues related to project implementation. (author)

  1. The Plant Cell, Vol. 11, 14451456, August 1999, www.plantcell.org 1999 American Society of Plant Physiologists Light QualityDependent Nuclear Import of the Plant

    E-Print Network [OSTI]

    Schäfer, Eberhard

    Physiologists Light Quality­Dependent Nuclear Import of the Plant Photoreceptors Phytochrome A and B StefanThe Plant Cell, Vol. 11, 1445­1456, August 1999, www.plantcell.org © 1999 American Society of Plant Institute of Plant Biology, Biological Research Center, P.O. Box 521, H-6701 Szeged, Hungary The phytochrome

  2. Next Generation Nuclear Plant Resilient Control System Functional Analysis

    SciTech Connect (OSTI)

    Lynne M. Stevens

    2010-07-01T23:59:59.000Z

    Control Systems and their associated instrumentation must meet reliability, availability, maintainability, and resiliency criteria in order for high temperature gas-cooled reactors (HTGRs) to be economically competitive. Research, perhaps requiring several years, may be needed to develop control systems to support plant availability and resiliency. This report functionally analyzes the gaps between traditional and resilient control systems as applicable to HTGRs, which includes the Next Generation Nuclear Plant; defines resilient controls; assesses the current state of both traditional and resilient control systems; and documents the functional gaps existing between these two controls approaches as applicable to HTGRs. This report supports the development of an overall strategy for applying resilient controls to HTGRs by showing that control systems with adequate levels of resilience perform at higher levels, respond more quickly to disturbances, increase operational efficiency, and increase public protection.

  3. Fiber optic sensors for nuclear power plant applications

    SciTech Connect (OSTI)

    Kasinathan, Murugesan; Sosamma, Samuel; BabuRao, Chelamchala; Murali, Nagarajan; Jayakumar, Tammana [Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu-603102 (India)

    2012-05-17T23:59:59.000Z

    Studies have been carried out for application of Raman Distributed Temperature Sensor (RDTS) in Nuclear Power Plants (NPP). The high temperature monitoring in sodium circuits of Fast Breeder Reactor (FBR) is important. It is demonstrated that RDTS can be usefully employed in monitoring sodium circuits and in tracking the percolating sodium in the surrounding insulation in case of any leak. Aluminum Conductor Steel Reinforced (ACSR) cable is commonly used as overhead power transmission cable in power grid. The suitability of RDTS for detecting defects in ACSR overhead power cable, is also demonstrated.

  4. Float level switch for a nuclear power plant containment vessel

    DOE Patents [OSTI]

    Powell, James G. (Clifton Park, NY)

    1993-01-01T23:59:59.000Z

    This invention is a float level switch used to sense rise or drop in water level in a containment vessel of a nuclear power plant during a loss of coolant accident. The essential components of the device are a guide tube, a reed switch inside the guide tube, a float containing a magnetic portion that activates a reed switch, and metal-sheathed, ceramic-insulated conductors connecting the reed switch to a monitoring system outside the containment vessel. Special materials and special sealing techniques prevent failure of components and allow the float level switch to be connected to a monitoring system outside the containment vessel.

  5. Float level switch for a nuclear power plant containment vessel

    DOE Patents [OSTI]

    Powell, J.G.

    1993-11-16T23:59:59.000Z

    This invention is a float level switch used to sense rise or drop in water level in a containment vessel of a nuclear power plant during a loss of coolant accident. The essential components of the device are a guide tube, a reed switch inside the guide tube, a float containing a magnetic portion that activates a reed switch, and metal-sheathed, ceramic-insulated conductors connecting the reed switch to a monitoring system outside the containment vessel. Special materials and special sealing techniques prevent failure of components and allow the float level switch to be connected to a monitoring system outside the containment vessel. 1 figures.

  6. Comparison of Options for a Pilot Plant Fusion Nuclear Mission

    SciTech Connect (OSTI)

    Brown, T; Goldston, R J; El-Guebaly, L; Kessel, C; Neilson, G H; Malang, S; Menard, J E; Prager, S; Waganer, L; Titus, P

    2012-08-27T23:59:59.000Z

    A fusion pilot plant study was initiated to clarify the development needs in moving from ITER to a first of a kind fusion power plant, following a path similar to the approach adopted for the commercialization of fission. The pilot plant mission encompassed component test and fusion nuclear science missions plus the requirement to produce net electricity with high availability in a device designed to be prototypical of the commercial device. Three magnetic configuration options were developed around this mission: the advanced tokamak (AT), spherical tokamak (ST) and compact stellarator (CS). With the completion of the study and separate documentation of each design option a question can now be posed; how do the different designs compare with each other as candidates for meeting the pilot plant mission? In a pro/con format this paper will examine the key arguments for and against the AT, ST and CS magnetic configurations. Key topics addressed include: plasma parameters, device configurations, size and weight comparisons, diagnostic issues, maintenance schemes, availability influences and possible test cell arrangement schemes.

  7. Nuclear power plant simulation facility evaluation methodology: handbook. Volume 1

    SciTech Connect (OSTI)

    Laughery, K.R. Jr.; Carter, R.J.; Haas, P.M.

    1986-01-01T23:59:59.000Z

    This report is Volume 1 of a two-part document which describes a project conducted to develop a methodology to evaluate the acceptability of nuclear power plant (NPP) simulation facilities for use in the simulator-based portion of NRC's operator licensing examination. The proposed methodology is to be utilized during two phases of the simulation facility life-cycle, initial simulator acceptance and recurrent analysis. The first phase is aimed at ensuring that the simulator provides an accurate representation of the reference NPP. There are two components of initial simulator evaluation: fidelity assessment and a direct determination of the simulation facility's adequacy for operator testing. The second phase is aimed at ensuring that the simulation facility continues to accurately represent the reference plant throughout the life of the simulator. Recurrent evaluation is comprised of three components: monitoring reference plant changes, monitoring the simulator's hardware, and examining the data from actual plant transients as they occur. Volume 1 is a set of guidelines which details the steps involved in the two life-cycle phases, presents an overview of the methodology and data collection requirements, and addresses the formation of the evaluation team and the preparation of the evaluation plan. 29 figs.

  8. Identification of good practices in the operation of nuclear power plants

    E-Print Network [OSTI]

    Chen, Haibo, 1975-

    2005-01-01T23:59:59.000Z

    This work developed an approach to diagnose problems and identify good practices in the operation of nuclear power plants using the system dynamics technique. The research began with construction of the ORSIM (Nuclear Power ...

  9. Incremental costs and optimization of in-core fuel management of nuclear power plants

    E-Print Network [OSTI]

    Watt, Hing Yan

    1973-01-01T23:59:59.000Z

    This thesis is concerned with development of methods for optimizing the energy production and refuelling decision for nuclear power plants in an electric utility system containing both nuclear and fossil-fuelled stations. ...

  10. An examination of the pursuit of nuclear power plant construction projects in the United States

    E-Print Network [OSTI]

    Guyer, Brittany (Brittany Leigh)

    2011-01-01T23:59:59.000Z

    The recent serious reconsideration of nuclear power as a means for U.S. electric utilities to increase their generation capacity provokes many questions regarding the achievable success of future nuclear power plant ...

  11. Study of seismic design bases and site conditions for nuclear power plants

    SciTech Connect (OSTI)

    Not Available

    1980-04-01T23:59:59.000Z

    This report presents the results of an investigation of four topics pertinent to the seismic design of nuclear power plants: Design accelerations by regions of the continental United States; review and compilation of design-basis seismic levels and soil conditions for existing nuclear power plants; regional distribution of shear wave velocity of foundation materials at nuclear power plant sites; and technical review of surface-founded seismic analysis versus embedded approaches.

  12. Microsoft Word - Table of Contents.docx

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Uncertainties for the Next Generation Nuclear Plant, INEELEXT-04-01816, June 2004. Engineering Services for the Next Generation Nuclear Plant (NGNP) with 750C Reactor...

  13. EIS-0225: Continued Operation of the Pantex Plant and Associated Storage of Nuclear Weapon Components

    Broader source: Energy.gov [DOE]

    This EIS evaluates the potential environemental impact of a proposal to continue operation of the Pantex Plant and associated storage of nuclear weapon components. Alternatives considered include: ...

  14. Aging of safety class 1E transformers in safety systems of nuclear power plants

    SciTech Connect (OSTI)

    Roberts, E.W.; Edson, J.L.; Udy, A.C. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1996-02-01T23:59:59.000Z

    This report discusses aging effects on safety-related power transformers in nuclear power plants. It also evaluates maintenance, testing, and monitoring practices with respect to their effectiveness in detecting and mitigating the effects of aging. The study follows the US Nuclear Regulatory Commission`s (NRC`s) Nuclear Plant-Aging Research approach. It investigates the materials used in transformer construction, identifies stressors and aging mechanisms, presents operating and testing experience with aging effects, analyzes transformer failure events reported in various databases, and evaluates maintenance practices. Databases maintained by the nuclear industry were analyzed to evaluate the effects of aging on the operation of nuclear power plants.

  15. Department of Mechanical and Nuclear Engineering Spring 2012 Automatic Plant Watering System

    E-Print Network [OSTI]

    Demirel, Melik C.

    PENNSTATE Department of Mechanical and Nuclear Engineering Spring 2012 Automatic Plant Watering System Overview The goal of this project was to design an automatic plant watering system for commercial in the soil of household plants and delivery water to those plants on a need-only basis. The overall design

  16. Feature Extraction for Data-Driven Fault Detection in Nuclear Power Plants Xin Jin, Robert M. Edwards and Asok Ray

    E-Print Network [OSTI]

    Ray, Asok

    Feature Extraction for Data-Driven Fault Detection in Nuclear Power Plants Xin Jin, Robert M monitoring of nuclear power plants (NPP) is one of the key issues addressed in nuclear energy safety research is performed during each nuclear power plant refueling outage, which may not be cost effective [1

  17. Aging of Class 1E batteries in safety systems of nuclear power plants

    SciTech Connect (OSTI)

    Edson, J.L.; Hardin, J.E.

    1987-07-01T23:59:59.000Z

    This report presents the results of a study of aging effects on safety-related batteries in nuclear power plants. The purpose is to evaluate the aging effects caused by operation within a nuclear facility and to evaluate maintenance, testing, and monitoring practices with respect to their effectiveness in detecting and mitigating the effects of aging. The study follows the US Nuclear Regulatory Commission's (NRC's) Nuclear Plant-Aging Research approach and investigates the materials used in battery construction, identifies stressors and aging mechanisms, presents operating and testing experience with aging effects, analyzes battery-failure events reported in various data bases, and evaluates recommended maintenance practices. Data bases that were analyzed included the NRC's Licensee Event Report system, the Institute for Nuclear Power Operations' Nuclear Plant Reliability Data System, the Oak Ridge National Laboratory's In-Plant Reliability Data System, and The S.M. Stoller Corporation's Nuclear Power Experience data base.

  18. System Analyses of High and Low-Temperature Interface Designs for a Nuclear-Driven High-Temperature Electrolysis Hydrogen Production Plant

    SciTech Connect (OSTI)

    E. A. Harvego; J. E. O'Brien

    2009-07-01T23:59:59.000Z

    As part of the Next Generation Nuclear Plant (NGNP) project, an evaluation of a low-temperature heat-pump interface design for a nuclear-driven high-temperature electrolysis (HTE) hydrogen production plant was performed using the UniSim process analysis software. The lowtemperature interface design is intended to reduce the interface temperature between the reactor power conversion system and the hydrogen production plant by extracting process heat from the low temperature portion of the power cycle rather than from the high-temperature portion of the cycle as is done with the current Idaho National Laboratory (INL) reference design. The intent of this design change is to mitigate the potential for tritium migration from the reactor core to the hydrogen plant, and reduce the potential for high temperature creep in the interface structures. The UniSim model assumed a 600 MWt Very-High Temperature Reactor (VHTR) operating at a primary system pressure of 7.0 MPa and a reactor outlet temperature of 900°C. The lowtemperature heat-pump loop is a water/steam loop that operates between 2.6 MPa and 5.0 MPa. The HTE hydrogen production loop operated at 5 MPa, with plant conditions optimized to maximize plant performance (i.e., 800°C electrolysis operating temperature, area specific resistance (ASR) = 0.4 ohm-cm2, and a current density of 0.25 amps/cm2). An air sweep gas system was used to remove oxygen from the anode side of the electrolyzer. Heat was also recovered from the hydrogen and oxygen product streams to maximize hydrogen production efficiencies. The results of the UniSim analysis showed that the low-temperature interface design was an effective heat-pump concept, transferring 31.5 MWt from the low-temperature leg of the gas turbine power cycle to the HTE process boiler, while consuming 16.0 MWe of compressor power. However, when this concept was compared with the current INL reference direct Brayton cycle design and with a modification of the reference design to simulate an indirect Brayton cycle (both with heat extracted from the high-temperature portion of the power cycle), the latter two concepts had higher overall hydrogen production rates and efficiencies compared to the low-temperature heatpump concept, but at the expense of higher interface temperatures. Therefore, the ultimate decision on the viability of the low-temperature heat-pump concept involves a tradeoff between the benefits of a lower-temperature interface between the power conversion system and the hydrogen production plant, and the reduced hydrogen production efficiency of the low-temperature heat-pump concept compared to concepts using high-temperature process heat.

  19. Corrosion and Creep of Candidate Alloys in High Temperature Helium and Steam Environments for the NGNP

    SciTech Connect (OSTI)

    Was, Gary; Jones, J. W.

    2013-06-21T23:59:59.000Z

    This project aims to understand the processes by which candidate materials degrade in He and supercritical water/steam environments characteristic of the current NGNP design. We will focus on understanding the roles of temperature, and carbon and oxygen potential in the 750-850 degree C range on both uniform oxidation and selective internal oxidation along grain boundaries in alloys 617 and 800H in supercritical water in the temperature range 500-600 degree C; and examining the application of static and cyclic stresses in combination with impure He environments in the temperature rang 750-850 degree C; and examining the application of static and cyclic stresses in combination with impure He environments in the temperature range 750-850 degree C over a range of oxygen and carbon potentials in helium. Combined, these studies wil elucidate the potential high damage rate processes in environments and alloys relevant to the NGNP.

  20. Management of aging of nuclear power plant containment structures

    SciTech Connect (OSTI)

    Naus, D.; Oland, C.B. [Oak Ridge National Lab., TN (United States). Engineering Technology Div.; Ellingwood, B. [Johns Hopkins Univ., Baltimore, MD (United States). Dept. of Civil Engineering; Norris, W.E.; Graves, H.L. III [Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research

    1998-06-01T23:59:59.000Z

    Research addressing aging management of nuclear power plant concrete and steel containment structures is summarized. Accomplishments related to concrete containment structures include formation of a materials` property database; an aging assessment methodology to identify critical structures and degradation factors; guidelines and evaluation criteria for use in condition assessments; and a time-dependent reliability-based methodology for condition assessments and estimations of future performance. Under the steel containments and liners activity, a degradation assessment methodology has been developed, mathematical models that describe time-dependent changes in the containment due to aggressive environmental factors have been identified, and statistical data supporting the use of these models in time-dependent reliability analysis have been summarized.

  1. Prognostics and Life Beyond 60 for Nuclear Power Plants

    SciTech Connect (OSTI)

    Leonard J. Bond; Pradeep Ramuhalli; Magdy S. Tawfik; Nancy J. Lybeck

    2011-06-01T23:59:59.000Z

    Safe, secure, reliable and sustainable energy supply is vital for advanced and industrialized life styles. To meet growing energy demand there is interest in longer term operation (LTO) for the existing nuclear power plant fleet and enhancing capabilities in new build. There is increasing use of condition based maintenance (CBM) for active components and periodic in service inspection (ISI) for passive systems: there is growing interest in deploying on-line monitoring. Opportunities exist to move beyond monitoring and diagnosis based on pattern recognition and anomaly detection to and prognostics with the ability to provide an estimate of remaining useful life (RUL). The adoption of digital I&C systems provides a framework within which added functionality including on-line monitoring can be deployed, and used to maintain and even potentially enhance safety, while at the same time improving planning and reducing both operations and maintenance costs.

  2. Progress Report for Diffusion Welding of the NGNP Process Application Heat Exchangers

    SciTech Connect (OSTI)

    R.E. Mizia; D.E. Clark; M.V. Glazoff; T.E. Lister; T.L. Trowbridge

    2011-04-01T23:59:59.000Z

    The NGNP Project is currently investigating the use of metallic, diffusion welded, compact heat exchangers to transfer heat from the primary (reactor side) heat transport system to the secondary heat transport system. The intermediate heat exchanger will transfer this heat to downstream applications such as hydrogen production, process heat, and electricity generation. The channeled plates that make up the heat transfer surfaces of the intermediate heat exchanger will have to be assembled into an array by diffusion welding.

  3. Underwater nuclear power plants: improved safety, environmental compatibility and efficiency

    SciTech Connect (OSTI)

    Galustov, K.Z.; Abadjyan, K.A.; Pavlov, A.B.

    1991-01-01T23:59:59.000Z

    The further development of nuclear power engineering depends on the creation of a new generation of nuclear power plant (NPP) projects that have a high degree of safety. Decisions ensuring secure NPP exploitation must be based on the possibility of eliminating or localizing accidents. Using environmental properties to achieve secure NPP exploitation and accident elimination leads to suggest the construction of NPPs in water. An efficient way to provide energy to remote coastal areas is through use of floatable construction of prefabricated units. Floatable construction raises the quality of works, reduces expenditures on industrial facilities, and facilities building conditions in districts with extreme climatic conditions. A type of NPP that is situated on a shelf with the reactor compartment placed at the sea bottom is proposed. The underwater location of the reactor compartment on the fixed depth allows the natural water environment conditions of natural hydrostatic pressure, heat transfer and circulation to provide NPP safety. An example of new concept for power units with under-water localization of the reactor compartment is provided by the double-block NPP in a VVER reactor.

  4. License Stewardship Approach to Commercial Nuclear Power Plant Decommissioning

    SciTech Connect (OSTI)

    Daly, P.T.; Hlopak, W.J. [Commercial Services Group, EnergySolutions 1009 Commerce Park, Oak Ridge, TN (United States)

    2008-07-01T23:59:59.000Z

    The paper explores both the conceptual approach to decommissioning commercial nuclear facilities using a license stewardship approach as well as the first commercial application of this approach. The license stewardship approach involves a decommissioning company taking control of a site and the 10 CFR 50 License in order to complete the work utilizing the established trust fund. In conclusion: The license stewardship approach is a novel way to approach the decommissioning of a retired nuclear power plant that offers several key advantages to all parties. For the owner and regulators, it provides assurance that the station will be decommissioned in a safe, timely manner. Ratepayers are assured that the work will be completed for the price they already have paid, with the decommissioning contractor assuming the financial risk of decommissioning. The contractor gains control of the assets and liabilities, the license, and the decommissioning fund. This enables the decommissioning contractor to control their work and eliminates redundant layers of management, while bringing more focus on achieving the desired end state - a restored site. (authors)

  5. Nuclear Safety Risk Management in Refueling Outage of Qinshan Nuclear Power Plant

    SciTech Connect (OSTI)

    Meijing Wu; Guozhang Shen [Qinshan Nuclear power company (China)

    2006-07-01T23:59:59.000Z

    The NPP is used to planning maintenance, in-service inspection, surveillance test, fuel handling and design modification in the refueling outage; the operator response capability will be reduced plus some of the plant systems out of service or loss of power at this time. Based on 8 times refueling outage experiences of the Qinshan NPP, this article provide some good practice and lesson learned for the nuclear safety risk management focus at four safety function areas of Residual Heat Removal Capability, Inventory Control, Power availability and Reactivity control. (authors)

  6. Pacific Basin Nuclear Conference (PBNC 2012), BEXCO, Busan, Korea, March 18 ~ 23, 2012 CHALLENGES OF CYBER SECURITY FOR NUCLEAR POWER PLANTS

    E-Print Network [OSTI]

    Kim, Kwangjo

    PBNC 2012 CHALLENGES OF CYBER SECURITY FOR NUCLEAR POWER PLANTS Kwangjo Kim KAIST, Daejeon, Korea.kim@kustar.ac.ae Abstract Nuclear Power Plants (NPPs) become one of the most important infrastructures in providing improvement. 1. Introduction Nuclear Power Plants (NPPs) become one of the most important infrastructures

  7. Contract Specifications For Olkiluoto 3 Nuclear Power Plant

    SciTech Connect (OSTI)

    Patrakka, Eero [Teollisuuden Voima Oy, 27160 Olkiluoto (Finland)

    2004-07-01T23:59:59.000Z

    The Finnish Parliament ratified in May 2002 the application for a Decision-in- Principle (DIP) that was submitted by Teollisuuden Voima Oy (TVO) in November 2000 concerning the construction of a new nuclear power plant in Finland (FIN5). The bid inquiries for FIN5 were sent out by TVO in September 2002, requesting the bids by the end of March 2003. A contract with the plant supplier was signed in December 2003, implying the construction of a PWR of type EPR (European Pressurised Water Reactor) in Olkiluoto, called Olkiluoto 3 NPP. The preparation of Bid Inquiry Specifications (BIS) was initiated simultaneously with the filing of the application for DIP. The compilation of BIS was an evolutionary process, starting with the collection of relevant reference material, proceeding through the development of technical, administrative and commercial requirements, and ending with the consolidation of all documentation to a package containing the complete BIS. An intensive bid evaluation process started immediately after receiving the bids, accompanied by negotiations with the supplier candidates. The final Contract Specifications (CS) were constituted on the basis of the BIS supplemented with information contained in the bid and the outcome of the contract negotiations. (author)

  8. A Literature Review on Ruthenium Behaviour in Nuclear Power Plant Severe Accidents

    E-Print Network [OSTI]

    Boyer, Edmond

    A Literature Review on Ruthenium Behaviour in Nuclear Power Plant Severe Accidents C. MUN , L Literature Review on Ruthenium Behaviour in Nuclear Power Plant Severe Accidents C. MUN a , L. CANTREL a , C Accidents Majeurs (DPAM), CEN Cadarache - France 1 b Commissariat à l'Energie Atomique (CEA), Direction de l'Energie

  9. The Decommissioning of the Trino Nuclear Power Plant

    SciTech Connect (OSTI)

    Brusa, L.; DeSantis, R.; Nurden, P. L.; Walkden, P.; Watson, B.

    2002-02-27T23:59:59.000Z

    Following a referendum in Italy in 1987, the four Nuclear Power Plants (NPPs) owned and operated by the state utility ENEL were closed. After closing the NPPs, ENEL selected a ''safestore'' decommissioning strategy; anticipating a safestore period of some 40-50 years. This approach was consistent with the funds collected during plant operation, and was reinforced by the lack of both a waste repository and a set of national free release limits for contaminated materials in Italy. During 1999, twin decisions were made to privatize ENEL and to transform the nuclear division into a separate subsidiary of the ENEL group. This group was renamed Sogin and during the following year, ownership of the company was transferred to the Italian Treasury. On formation, Sogin was asked by the Italian government to review the national decommissioning strategy. The objective of the review was to move from a safestore strategy to a prompt decommissioning strategy, with the target of releasing all of the nuclear sites by 2020. It was recognized that this target was conditional upon the availability of a national LLW repository together with interim stores for both spent fuel and HLW by 2009. The government also agreed that additional costs caused by the acceleration of the decommissioning program would be considered as stranded costs. These costs will be recovered by a levy on the kWh price of electricity, a process established and controlled by the Regulator of the Italian energy sector. Building on the successful collaboration to develop a prompt decommissioning strategy for the Latina Magnox reactor (1), BNFL and Sogin agreed to collaborate on an in depth study for the prompt decommissioning of the Sogin PWR at Trino. BNFL is currently decommissioning six NPPs and is at an advanced stage of planning for two further units, having completed a full and rigorous exercise to develop Baseline Decommissioning Plans (BDP's) for these stations. The BDP exercise utilizes the full range of BNFL decommissioning experience and knowledge to develop a strategy, methodology and cost for the decommissioning of NPPs. Over the past year, a prompt decommissioning strategy for Trino has been developed. The strategy has been based on the principles of minimizing waste products that require long term storage, maximizing 'free release' materials and utilizing existing and regulatory approved technologies.

  10. Leasing of Nuclear Power Plants With Using Floating Technologies

    SciTech Connect (OSTI)

    Kuznetsov, Yu.N.; Gabaraev, B.A.; Reshetov, V.A.; Moskin, V.A. [Federal State Unitary Enterprise, N.A. Dollezhal' Scientific-Research and Design Institute of Power Engineering (Russian Federation)

    2002-07-01T23:59:59.000Z

    The proposal to organize and realize the international program on leasing of Nuclear Power Plant (NPP) reactor compartments is brought to the notice of potential partners. The proposal is oriented to the construction of new NPPs or to replacement of worked-out reactor units of the NPPs in operation on the sites situated near water area and to the use of afloat technologies for construction, mounting and transportation of reactor units as a Reactor Compartment Block Module (RCBM). According to the offered project the RCBM is fabricated in factory conditions at the largest Russian defense shipbuilding plant - State Unitary Enterprise 'Industrial Association SEVMASHPREDPRIYATIE' (SEVMASH) in the city of Severodvinsk of the Arkhangelsk region. After completion of assembling, testing and preliminary licensing the RCBM is given buoyancy by means of hermetic sealing and using pontoons and barges. The RCBM delivery to the NPP site situated near water area is performed by sea route. The RCBM is brought to the place of its installation with the use of appropriate hydraulic structures (canals, shipping locks), then is lowered on the basement constructed beforehand and incorporated into NPP scheme, of which the components are installed in advance. Floating means can be detached from the RCBM and used repeatedly for other RCBMs. Further procedure of NPP commissioning and its operation is carried out according to traditional method by power company in the framework of RCBM leasing with enlisting the services of firm-manufacturer's specialists either to provide reactor plant operation and concomitant processes or to perform author's supervision of operation. After completion of lifetime and reactor unloading the RCBM is dismantled with using the same afloat technology and taken away from NPP site to sea area entirely, together with its structures (reactor vessel, heat exchangers, pumps, pipelines and other equipment). Then RCBM is transported by shipping route to a firm-manufacturer, for subsequent reprocessing, utilization and storage. Nuclear fuel and radioactive wastes are removed from NPP site also. Use of leasing method removes legal problems connected with the transportation of radioactive materials through state borders as the RCBM remains a property of the state-producer at all stages of its life cycle. (authors)

  11. Feasibility Study of Hydrogen Production at Existing Nuclear Power Plants

    SciTech Connect (OSTI)

    Stephen Schey

    2009-07-01T23:59:59.000Z

    Cooperative Agreement DE-FC07-06ID14788 was executed between the U.S. Department of Energy, Electric Transportation Applications, and Idaho National Laboratory to investigate the economics of producing hydrogen by electrolysis using electricity generated by nuclear power. The work under this agreement is divided into the following four tasks: Task 1 – Produce Data and Analyses Task 2 – Economic Analysis of Large-Scale Alkaline Electrolysis Task 3 – Commercial-Scale Hydrogen Production Task 4 – Disseminate Data and Analyses. Reports exist on the prospect that utility companies may benefit from having the option to produce electricity or produce hydrogen, depending on market conditions for both. This study advances that discussion in the affirmative by providing data and suggesting further areas of study. While some reports have identified issues related to licensing hydrogen plants with nuclear plants, this study provides more specifics and could be a resource guide for further study and clarifications. At the same time, this report identifies other area of risks and uncertainties associated with hydrogen production on this scale. Suggestions for further study in some of these topics, including water availability, are included in the report. The goals and objectives of the original project description have been met. Lack of industry design for proton exchange membrane electrolysis hydrogen production facilities of this magnitude was a roadblock for a significant period. However, recent design breakthroughs have made costing this facility much more accurate. In fact, the new design information on proton exchange membrane electrolyzers scaled to the 1 kg of hydrogen per second electrolyzer reduced the model costs from $500 to $100 million. Task 1 was delayed when the original electrolyzer failed at the end of its economic life. However, additional valuable information was obtained when the new electrolyzer was installed. Products developed during this study include a process model and a N2H2 economic assessment model (both developed by the Idaho National Laboratory). Both models are described in this report. The N2H2 model closely tracked and provided similar results as the H2A model and was instrumental in assessing the effects of plant availability on price when operated in the shoulder mode for electrical pricing. Differences between the H2A and N2H2 model are included in this report.

  12. Thirty states sign ITER nuclear fusion plant deal 1 hour, 28 minutes ago

    E-Print Network [OSTI]

    Thirty states sign ITER nuclear fusion plant deal 1 hour, 28 minutes ago Representatives of more than 30 countries signed a deal on Tuesday to build the world's most advanced nuclear fusion reactor nuclear reactors, but critics argue it could be at least 50 years before a commercially viable reactor

  13. POWER PLANT RELIABILITY-AVAILABILITY AND STATE REGULATION. VOLUME 7 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01T23:59:59.000Z

    Removal Equipment (nuclear plant) Turbine Building ClosedCooling Water System (nuclear plant) SteamReheater (nuclear plant) Inspection Water Induction

  14. The impact of offsite factors on the safety performance of small nuclear power plants

    SciTech Connect (OSTI)

    Baranaev, Yu.D.; Viktorov, A.N. [Inst. of Physics and Power Engineering, Obninsk (Russian Federation)

    1991-01-01T23:59:59.000Z

    The results of an analysis of the influence of offsite factors on small nuclear power-plant (SNPP) safety performance during postulated severe accidents are presented. Given the plant locations in the immediate vicinity of residential areas and the impossibility of accomplishing the expeditious evacuation of the public, the risk caused by an SNPP severe accident may be considerably less than that for such an event in a large nuclear power plant. 3 refs., 3 figs., 5 tabs.

  15. CONTROL OF POPULATION DENSITIES SURROUNDING NUCLEAR POWER PLANTS. VOLUME 5 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, jA.V.

    2010-01-01T23:59:59.000Z

    Related Standards for Nuclear Power Plants," by A.V. NeroResponse Planning for Nuclear Power Plants in California,"Densities Surrounding Nuclear Power Plants," by A.V. Nero,

  16. CONTROL OF POPULATION DENSITIES SURROUNDING NUCLEAR POWER PLANTS. VOLUME 5 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, jA.V.

    2010-01-01T23:59:59.000Z

    Standards for Nuclear Power Plants," by A.V. Nero and Y.C.Planning for Nuclear Power Plants in California," by W.W.S.Surrounding Nuclear Power Plants," by A.V. Nero, C.H.

  17. Example G Cost of construction of nuclear power plants Description of data

    E-Print Network [OSTI]

    Reid, Nancy

    1 Example G Cost of construction of nuclear power plants Description of data Table G.1 gives reactor (LWR) power plants constructed in USA. It is required to predict the capital cost involved in the construction of further LWR power plants. The notation used in Table G.1 is explained in Table G.2. The final 6

  18. Example G Cost of construction of nuclear power plants Description of data

    E-Print Network [OSTI]

    Reid, Nancy

    Example G Cost of construction of nuclear power plants Description of data Table G.1 gives data) power plants constructed in USA. It is required to predict the capital cost involved in the construction of further LWR power plants. The notation used in Table G.1 is explained in Table G.2. The final 6 lines

  19. Department of Mechanical and Nuclear Engineering Spring 2012 East Campus Power Plant Deaerator Optimization

    E-Print Network [OSTI]

    Demirel, Melik C.

    PENNSTATE Department of Mechanical and Nuclear Engineering Spring 2012 East Campus Power Plant Deaerator Optimization Overview In the East Campus Power plant a new Deaerator system has been installed which Deaerator is the most efficient and then make a recommendation to the plant of which one should

  20. Waste Minimization Policy at the Romanian Nuclear Power Plant

    SciTech Connect (OSTI)

    Andrei, V.; Daian, I.

    2002-02-26T23:59:59.000Z

    The radioactive waste management system at Cernavoda Nuclear Power Plant (NPP) in Romania was designed to maintain acceptable levels of safety for workers and to protect human health and the environment from exposure to unacceptable levels of radiation. In accordance with terminology of the International Atomic Energy Agency (IAEA), this system consists of the ''pretreatment'' of solid and organic liquid radioactive waste, which may include part or all of the following activities: collection, handling, volume reduction (by an in-drum compactor, if appropriate), and storage. Gaseous and aqueous liquid wastes are managed according to the ''dilute and discharge'' strategy. Taking into account the fact that treatment/conditioning and disposal technologies are still not established, waste minimization at the source is a priority environmental management objective, while waste minimization at the disposal stage is presently just a theoretical requirement for future adopted technologies . The necessary operational and maintenance procedures are in place at Cernavoda to minimize the production and contamination of waste. Administrative and technical measures are established to minimize waste volumes. Thus, an annual environmental target of a maximum 30 m3 of radioactive waste volume arising from operation and maintenance has been established. Within the first five years of operations at Cernavoda NPP, this target has been met. The successful implementation of the waste minimization policy has been accompanied by a cost reduction while the occupational doses for plant workers have been maintained at as low as reasonably practicable levels. This paper will describe key features of the waste management system along with the actual experience that has been realized with respect to minimizing the waste volumes at the Cernavoda NPP.

  1. Nuclear Energy Research Initiative. Risk Informed Assessment of Regulatory and Design Requirements for Future Nuclear Power Plants. Annual Report

    SciTech Connect (OSTI)

    Ritterbusch, S.E.

    2000-08-01T23:59:59.000Z

    The overall goal of this research project is to support innovation in new nuclear power plant designs. This project is examining the implications, for future reactors and future safety regulation, of utilizing a new risk-informed regulatory system as a replacement for the current system. This innovation will be made possible through development of a scientific, highly risk-informed approach for the design and regulation of nuclear power plants. This approach will include the development and.lor confirmation of corresponding regulatory requirements and industry standards. The major impediment to long term competitiveness of new nuclear plants in the U.S. is the capital cost component--which may need to be reduced on the order of 35% to 40% for Advanced Light Water Reactors (ALWRs) such as System 80+ and Advanced Boiling Water Reactor (ABWR). The required cost reduction for an ALWR such as AP600 or AP1000 would be expected to be less. Such reductions in capital cost will require a fundamental reevaluation of the industry standards and regulatory bases under which nuclear plants are designed and licensed. Fortunately, there is now an increasing awareness that many of the existing regulatory requirements and industry standards are not significantly contributing to safety and reliability and, therefore, are unnecessarily adding to nuclear plant costs. Not only does this degrade the economic competitiveness of nuclear energy, it results in unnecessary costs to the American electricity consumer. While addressing these concerns, this research project will be coordinated with current efforts of industry and NRC to develop risk-informed, performance-based regulations that affect the operation of the existing nuclear plants; however, this project will go farther by focusing on the design of new plants.

  2. Safeguards Issues at Nuclear Reactors and Enrichment Plants

    SciTech Connect (OSTI)

    Boyer, Brian D [Los Alamos National Laboratory

    2012-08-15T23:59:59.000Z

    The Agency's safeguards technical objective is the timely detection of diversion of significant quantities of nuclear material from peaceful nuclear activities to the manufacture of nuclear weapons or of other nuclear explosive devices or for purposes unknown, and deterrence of such diversion by the risk of early detection.

  3. Two novel procedures for aggregating randomized model ensemble outcomes for robust signal reconstruction in nuclear power plants monitoring systems

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    reconstruction in nuclear power plants monitoring systems P. Baraldi1 , E. Zio1,* , G. Gola2 , D. Roverso2 , M importance for the safe and reliable operation of nuclear power plants. Auto-associative regression models of nuclear power plants for it allows the timely detection of malfunctions and anomalies during operation

  4. Abstract--Resins are used in nuclear power plants for water ultrapurification. Two approaches are considered in this work

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Abstract--Resins are used in nuclear power plants for water ultrapurification. Two approaches in manufacturing ultrapure water for nuclear power plants. Resins allow the removal of ionic impurities to subparts-per-million. Thereby in nuclear power plants, resins contribute to guarantee personnel safety, to control feed system

  5. DATA-DRIVEN ON-LINE PREDICTION OF THE AVAILABLE RECOVERY TIME IN NUCLEAR POWER PLANT FAILURE SCENARIOS

    E-Print Network [OSTI]

    Boyer, Edmond

    1 DATA-DRIVEN ON-LINE PREDICTION OF THE AVAILABLE RECOVERY TIME IN NUCLEAR POWER PLANT FAILURE-XADS). Key Words: Recovery Time, Emergency Accident Management, Nuclear Power Plant, Lead- Bismuth Eutectic e [Øwre, 2001]. Yet, the problem of what kind of decision support to provide to nuclear power plant

  6. Conference committees Chairman of the conference "New Nuclear Power Plants in the Netherlands", June 21-22, 2011,

    E-Print Network [OSTI]

    Conference committees Chairman of the conference "New Nuclear Power Plants in the Netherlands Nuclear Power Plants, September 15-19, 2003, Kyoto, Japan. Session chairman GENES4/ANP2003 ,,International Conference on Global Environment and Advanced Nuclear Power Plants, September 15-19, 2003, Kyoto

  7. Status of radioiodine control for nuclear fuel reprocessing plants

    SciTech Connect (OSTI)

    Burger, L.L.; Scheele, R.D.

    1983-07-01T23:59:59.000Z

    This report summarizes the status of radioiodine control in a nuclear fuel reprocessing plant with respect to capture, fixation, and disposal. Where possible, we refer the reader to a number of survey documents which have been published in the last four years. We provide updates where necessary. Also discussed are factors which must be considered in developing criteria for iodine control. For capture from gas streams, silver mordenite and a silver nitrate impregnated silica (AC-6120) are considered state-of-the-art and are recommended. Three aqueous scrubbing processes have been demonstrated: Caustic scrubbing is simple but probably will not give an adequate iodine retention by itself. Mercurex (mercuric nitrate-nitric acid scrubbing) has a number of disadvantages including the use of toxic mercury. Iodox (hyperazeotropic nitric acid scrubbing) is effective but employs a very corrosive and hazardous material. Other technologies have been tested but require extensive development. The waste forms recommended for long-term storage or disposal are silver iodide, the iodates of barium, strontium, or calcium, and silver loaded sorbents, all fixed in cement. Copper iodide in bitumen (asphalt) is a possibility but requires testing. The selection of a specific form will be influenced by the capture process used.

  8. Reassessment of selected factors affecting siting of Nuclear Power Plants

    SciTech Connect (OSTI)

    Davis, R.E.; Hanson, A.L.; Mubayi, V.; Nourbakhsh, H.P.

    1997-02-01T23:59:59.000Z

    Brookhaven National Laboratory has performed a series of probabilistic consequence assessment calculations for nuclear reactor siting. This study takes into account recent insights into severe accident source terms and examines consequences in a risk based format consistent with the quantitative health objectives (QHOs) of the NRC`s Safety Goal Policy. Simplified severe accident source terms developed in this study are based on the risk insights of NUREG-1150. The results of the study indicate that both the quantity of radioactivity released in a severe accident as well as the likelihood of a release are lower than those predicted in earlier studies. The accident risks using the simplified source terms are examined at a series of generic plant sites, that vary in population distribution, meteorological conditions, and exclusion area boundary distances. Sensitivity calculations are performed to evaluate the effects of emergency protective action assumptions on the risk of prompt fatality and latent cancers fatality, and population relocation. The study finds that based on the new source terms the prompt and latent fatality risks at all generic sites meet the QHOs of the NRC`s Safety Goal Policy by margins ranging from one to more than three orders of magnitude. 4 refs., 17 figs., 24 tabs.

  9. Aging assessment of surge protective devices in nuclear power plants

    SciTech Connect (OSTI)

    Davis, J.F.; Subudhi, M. [Brookhaven National Lab., Upton, NY (United States)] [Brookhaven National Lab., Upton, NY (United States); Carroll, D.P. [Florida Univ., Gainesville, FL (United States)] [Florida Univ., Gainesville, FL (United States)

    1996-01-01T23:59:59.000Z

    An assessment was performed to determine the effects of aging on the performance and availability of surge protective devices (SPDs), used in electrical power and control systems in nuclear power plants. Although SPDs have not been classified as safety-related, they are risk-important because they can minimize the initiating event frequencies associated with loss of offsite power and reactor trips. Conversely, their failure due to age might cause some of those initiating events, e.g., through short circuit failure modes, or by allowing deterioration of the safety-related component(s) they are protecting from overvoltages, perhaps preventing a reactor trip, from an open circuit failure mode. From the data evaluated during 1980--1994, it was found that failures of surge arresters and suppressers by short circuits were neither a significant risk nor safety concern, and there were no failures of surge suppressers preventing a reactor trip. Simulations, using the ElectroMagnetic Transients Program (EMTP) were performed to determine the adequacy of high voltage surge arresters.

  10. Quiz # 7, STAT 383, Prof. Suman Sanyal, April 8, 2009 (Q2, Page 354) To decide whether the pipe welds in a nuclear power plant meet

    E-Print Network [OSTI]

    Sanyal, Suman

    welds in a nuclear power plant meet specifications, a random sample of welds is to be selected : µ nuclear power plants is to determine if welds

  11. Confirmation of the seismic resistance of nuclear power plant equipment after assembly

    SciTech Connect (OSTI)

    Kaznovsky, P. S.; Kaznovsky, A. P.; Saakov, E. S.; Ryasnyj, S. I. [JSC 'Atomtehenergo' (Russian Federation)

    2013-05-15T23:59:59.000Z

    It is shown that the natural frequencies and damping decrements of nuclear power plant equipment can only be determined experimentally and directly at the power generation units (reactors) of nuclear power plants under real disassembly conditions for the equipment, piping network, thermal insulation, etc. A computational experimental method is described in which the natural frequencies and damping decrements are determined in the field and the seismic resistance is reevaluated using these values. This method is the basis of the standards document 'Methods for confirming the dynamic characteristics of systems and components of the generating units of nuclear power plants which are important for safety' prepared and introduced in 2012.

  12. Atmospheric dispersion and the radiological consequences of normal airborne effluents from a nuclear power plant

    SciTech Connect (OSTI)

    Fang, D.; Yang, L. [Tsinghua Univ., Beijing (China); Sun, C.Z. [Suhou Nuclear Research Inst., Suzhou (China)

    1995-01-01T23:59:59.000Z

    The relationship between the consequences of the normal exhaust of radioactive materials in air from nuclear power plants and atmospheric dispersion is studied. Because the source terms of the exhaust from a nuclear power plant are relatively low and their radiological consequences are far less than the corresponding authoritative limits, the atmospheric dispersion models, their various modifications, and selections of relevant parameters have few effects on those consequences. In the environmental assessment and siting, the emphasis should not be placed on the consequence evaluation of routine exhaust of nuclear power plants, and the calculation of consequences of the exhaust and atmospheric field measurements should be appropriately, simplified. 12 refs., 5 figs., 7 tabs.

  13. NE Press Releases | Department of Energy

    Broader source: Energy.gov (indexed) [DOE]

    For Irradiation Performance As part of the Office of Nuclear Energy's Next Generation Nuclear Plant (NGNP) Program, the Advanced Gas Reactor (AGR) Fuel Development Program has...

  14. Microsoft Word - 911146_0_SSC-15 Primary Circuit and BOP Inst...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    6 Revision 0 ENGINEERING SERVICES FOR THE NEXT GENERATION NUCLEAR PLANT (NGNP) WITH HYDROGEN PRODUCTION Test Plan for Primary Circuit and Balance of Plant Instrumentation Prepared...

  15. Design Configurations and Coupling High Temperature Gas-Cooled Reactor and Hydrogen Plant

    SciTech Connect (OSTI)

    Chang H. Oh; Eung Soo Kim; Steven Sherman

    2008-04-01T23:59:59.000Z

    The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood.

  16. Probabilistic methods in seismic risk assessment for nuclear power plants: proceedings

    SciTech Connect (OSTI)

    Not Available

    1983-01-01T23:59:59.000Z

    The state-of-the-art in seismic risk analysis applied to the design and siting of nuclear power plants was addressed in this meeting. Presentations were entered individually into the date base. (ACR)

  17. Maximizing nuclear power plant performance via mega-uprates and subsequent license renewal

    E-Print Network [OSTI]

    DeWitte, Jacob D. (Jacob Dominic)

    2014-01-01T23:59:59.000Z

    The goal of this thesis is to develop a methodology to evaluate the engineering and economic implications of maximizing performance of the United States' commercial fleet of nuclear power plants. This methodology addresses ...

  18. Dynamic reliability using entry-time approach for maintenance of nuclear power plants

    E-Print Network [OSTI]

    Wang, Shuwen

    2009-05-15T23:59:59.000Z

    -time processes have the potential to provide a significantly greater range of applicability and flexibility than traditional reliability tools for case studies related to equipment and components in nuclear power plants. In this dissertation, the finite...

  19. A holistic investigation of complexity sources in nuclear power plant control rooms

    E-Print Network [OSTI]

    Sasangohar, Farzan

    2011-01-01T23:59:59.000Z

    The nuclear power community in the United States is moving to modernize aging power plant control rooms as well as develop control rooms for new reactors. New generation control rooms, along with modernized control rooms, ...

  20. Nuclear Plant Feedwater Heater Handbook. Volume 3. Operation and maintenance guidelines. Final report

    SciTech Connect (OSTI)

    Bell, R.J.; Hardy, C.D. Jr.

    1985-06-01T23:59:59.000Z

    This document is the third part of a three-volume handbook covering closed feedwater heaters for nuclear electric power generating plants. This third volume covers the operation and maintenance of closed feedwater heaters. 11 refs., 23 figs., 5 tabs.

  1. Nuclear Plant Feedwater Heater Handbook. Volume 1. Primer. Final report

    SciTech Connect (OSTI)

    Bell, R.J.; Wells, T.G. Jr.

    1985-06-01T23:59:59.000Z

    This document is the first part of a three volume handbook covering closed feedwater heaters for electric power generating plants. This volume is a primer to the subject of feedwater heaters and their integration into the plant. 24 refs.

  2. Integration of Biorefineries and Nuclear Cogeneration Power Plants - A Preliminary Analysis

    SciTech Connect (OSTI)

    Greene, Sherrell R [ORNL; Flanagan, George F [ORNL; Borole, Abhijeet P [ORNL

    2009-03-01T23:59:59.000Z

    Biomass-based ethanol and nuclear power are two viable elements in the path to U.S. energy independence. Numerous studies suggest nuclear power could provide a practical carbon-free heat source alternative for the production of biomass-based ethanol. In order for this coupling to occur, it is necessary to examine the interfacial requirements of both nuclear power plants and bioethanol refineries. This report describes the proposed characteristics of a small cogeneration nuclear power plant, a biochemical process-based cellulosic bioethanol refinery, and a thermochemical process-based cellulosic biorefinery. Systemic and interfacial issues relating to the co-location of either type of bioethanol facility with a nuclear power plant are presented and discussed. Results indicate future co-location efforts will require a new optimized energy strategy focused on overcoming the interfacial challenges identified in the report.

  3. Risk-informed assessment of regulatory and design requirements for future nuclear power plants. Annual report

    SciTech Connect (OSTI)

    NONE

    2000-08-01T23:59:59.000Z

    OAK B188 Risk-informed assessment of regulatory and design requirements for future nuclear power plants. Annual report. The overall goal of this research project is to support innovation in new nuclear power plant designs. This project is examining the implications, for future reactors and future safety regulation, of utilizing a new risk-informed regulatory system as a replacement for the current system. This innovation will be made possible through development of a scientific, highly risk-formed approach for the design and regulation of nuclear power plants. This approach will include the development and/or confirmation of corresponding regulatory requirements and industry standards. The major impediment to long term competitiveness of new nuclear plants in the U.S. is the capital cost component--which may need to be reduced on the order of 35% to 40% for Advanced Light Water Reactors (ALWRS) such as System 80+ and Advanced Boiling Water Reactor (ABWR). The required cost reduction for an ALWR such as AP600 or AP1000 would be expected to be less. Such reductions in capital cost will require a fundamental reevaluation of the industry standards and regulatory bases under which nuclear plants are designed and licensed. Fortunately, there is now an increasing awareness that many of the existing regulatory requirements and industry standards are not significantly contributing to safety and reliability and, therefore, are unnecessarily adding to nuclear plant costs. Not only does this degrade the economic competitiveness of nuclear energy, it results in unnecessary costs to the American electricity consumer. While addressing these concerns, this research project will be coordinated with current efforts of industry and NRC to develop risk-informed, performance-based regulations that affect the operation of the existing nuclear plants; however, this project will go further by focusing on the design of new plants.

  4. Potential safety-related incidents with possible applicability to a nuclear fuel reprocessing plant

    SciTech Connect (OSTI)

    Perkins, W.C.; Durant, W.S.; Dexter, A.H.

    1980-12-01T23:59:59.000Z

    The occurrence of certain potential events in nuclear fuel reprocessing plants could lead to significant consequences involving risk to operating personnel or to the general public. This document is a compilation of such potential initiating events in nuclear fuel reprocessing plants. Possible general incidents and incidents specific to key operations in fuel reprocessing are considered, including possible causes, consequences, and safety features designed to prevent, detect, or mitigate such incidents.

  5. Basis for NGNP Reactor Design Down-Selection

    SciTech Connect (OSTI)

    L.E. Demick

    2010-08-01T23:59:59.000Z

    The purpose of this paper is to identify the extent of technology development, design and licensing maturity anticipated to be required to credibly identify differences that could make a technical choice practical between the prismatic and pebble bed reactor designs. This paper does not address a business decision based on the economics, business model and resulting business case since these will vary based on the reactor application. The selection of the type of reactor, the module ratings, the number of modules, the configuration of the balance of plant and other design selections will be made on the basis of optimizing the Business Case for the application. These are not decisions that can be made on a generic basis.

  6. Basis for NGNP Reactor Design Down-Selection

    SciTech Connect (OSTI)

    L.E. Demick

    2011-11-01T23:59:59.000Z

    The purpose of this paper is to identify the extent of technology development, design and licensing maturity anticipated to be required to credibly identify differences that could make a technical choice practical between the prismatic and pebble bed reactor designs. This paper does not address a business decision based on the economics, business model and resulting business case since these will vary based on the reactor application. The selection of the type of reactor, the module ratings, the number of modules, the configuration of the balance of plant and other design selections will be made on the basis of optimizing the Business Case for the application. These are not decisions that can be made on a generic basis.

  7. A Domain-Specific Safety Analysis for Digital Nuclear Plant Protection Systems

    E-Print Network [OSTI]

    A Domain-Specific Safety Analysis for Digital Nuclear Plant Protection Systems Sanghyun Yoon tries to assure the systems' safety through performing various safety analysis techniques ­ FTA (Fault was KNICS(Korea Nuclear Instrumentation and Control System) RPS(Reactor Protection System). · Prototype

  8. Risk Informed Assessment of Regulatory and Design Requirements for Future Nuclear Power Plants - Final Technical Report

    SciTech Connect (OSTI)

    Ritterbusch, Stanley; Golay, Michael; Duran, Felicia; Galyean, William; Gupta, Abhinav; Dimitrijevic, Vesna; Malsch, Marty

    2003-01-29T23:59:59.000Z

    OAK B188 Summary of methods proposed for risk informing the design and regulation of future nuclear power plants. All elements of the historical design and regulation process are preserved, but the methods proposed for new plants use probabilistic risk assessment methods as the primary decision making tool.

  9. Assessement of Codes and Standards Applicable to a Hydrogen Production Plant Coupled to a Nuclear Reactor

    SciTech Connect (OSTI)

    M. J. Russell

    2006-06-01T23:59:59.000Z

    This is an assessment of codes and standards applicable to a hydrogen production plant to be coupled to a nuclear reactor. The result of the assessment is a list of codes and standards that are expected to be applicable to the plant during its design and construction.

  10. Nuclear material safeguards for enrichments plants: Part 4, Gas Centrifuge Enrichment Plant: Diversion scenarios and IAEA safeguards activities: Safeguards training course

    SciTech Connect (OSTI)

    Not Available

    1988-10-01T23:59:59.000Z

    This publication is Part 4 of a safeguards training course in Nuclear Material Safeguards for enrichment plants. This part of the course deals with diversion scenarios and safeguards activities at gas centrifuge enrichment plants.

  11. In-Plant Reliability Data base for nuclear plant components. Interim report: diesel generators, batteries, chargers and inverters

    SciTech Connect (OSTI)

    Kahl, W.K.; Borkowski, R.J.

    1985-01-01T23:59:59.000Z

    The objective of the In-Plant Reliability Data (IPRD) program is to develop a comprehensive, component-specific reliability data base for probabilistic risk assessment and for other statistical analyses relevant to component reliability evaluations. This document is the product of a pilot study that was undertaken to demonstrate the methodology and feasibility of applying IPRDS techniques to develop and analyze the reliability characteristics of key electrical components in five nuclear power plants. These electrical components include diesel generators, batteries, battery chargers and inverters. The sources used to develop the data base and produce the component failure rates and mean repair times were the plant equipment lists, plant drawings, maintenance work requests, Final Safety Analysis Reports (FSARs), and interviews with plant personnel. The data spanned approximately 33 reactor-years of commercial operation.

  12. Prognostics Health Management and Life Beyond 60 for Nuclear Power Plants

    SciTech Connect (OSTI)

    Ramuhalli, Pradeep; Coble, Jamie B.; Meyer, Ryan M.; Bond, Leonard J.

    2013-12-01T23:59:59.000Z

    There is growing interest in longer-term operation of the current US nuclear power plant fleet. This paper will present an overview of prognostic health management (PHM) technologies that could play a role in the safe and effective operation of nuclear power plants during extended life. A case study in prognostics for materials degradation assessment, using laboratory-scale measurements, is briefly discussed, and technical gaps that need to be addressed prior to PHM system deployment for nuclear power life extension are presented.

  13. advanced nuclear plant: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  14. advanced nuclear plants: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  15. Sensitivity analysis for the outages of nuclear power plants

    E-Print Network [OSTI]

    2012-02-17T23:59:59.000Z

    Feb 17, 2012 ... Energy generation in France is a competitive market, whereas ... from wind farms, solar energy or run of river plant without pondage.

  16. Dynamic alarm presentation in a nuclear plant control room

    DOE Patents [OSTI]

    Kenneth, Scarola (Windsor, CT); Jamison, David S. (Windsor, CT); Manazir, Richard M. (North Canton, CT); Rescorl, Robert L. (Vernon, CT); Harmon, Daryl L. (Enfield, CT)

    1994-01-01T23:59:59.000Z

    The alarm activation set point and priority for a given, spatially fixed alarm tile can vary depending in part on the mode of plant operation.

  17. Incentive regulation of investor-owned nuclear power plants by public utility regulators. Revision 1

    SciTech Connect (OSTI)

    McKinney, M.D.; Seely, H.E.; Merritt, C.R.; Baker, D.C. [Pacific Northwest Lab., Richland, WA (United States)

    1995-04-01T23:59:59.000Z

    The US Nuclear Regulatory Commission (NRC) periodically surveys the Federal Energy Regulatory Commission (FERC) and state regulatory commissions that regulate utility owners of nuclear power plants. The NRC is interested in identifying states that have established economic or performance incentive programs applicable to nuclear power plants, how the programs are being implemented, and in determining the financial impact of the programs on the utilities. The NRC interest stems from the fact that such programs have the potential to adversely affect the safety of nuclear power plants. The current report is an update of NUREG/CR-5975, Incentive Regulation of Investor-Owned Nuclear Power Plants by Public Utility Regulators, published in January 1993. The information in this report was obtained from interviews conducted with each state regulatory agency that administers an incentive program and each utility that owns at least 10% of an affected nuclear power plant. The agreements, orders, and settlements that form the basis for each incentive program were reviewed as required. The interviews and supporting documentation form the basis for the individual state reports describing the structure and financial impact of each incentive program.

  18. Nuclear norm minimization for the planted clique and biclique ...

    E-Print Network [OSTI]

    2009-01-21T23:59:59.000Z

    Jan 21, 2009 ... This problem was shown to be. NP-hard by Peeters [16]. In Sections 3 and 4, we relax these problems to convex optimization using the nuclear.

  19. Nuclear Power Plant NDE Challenges - Past, Present, and Future

    SciTech Connect (OSTI)

    Doctor, Steven R.

    2007-01-01T23:59:59.000Z

    This is a paper that covers the major thrust of NDE work that PNNL has conducted for the U.S. Nuclear Regulatory Commission from 1977 to the present.

  20. Identification of performance indicators for nuclear power plants

    E-Print Network [OSTI]

    Sui, Yu, 1973-

    2001-01-01T23:59:59.000Z

    Performance indicators have been assuming an increasingly important role in the nuclear industry. An integrated methodology is proposed in this research for the identification and validation of performance indicators for ...

  1. Source book for planning nuclear dual-purpose electric/distillation desalination plants

    SciTech Connect (OSTI)

    Reed, S.A.

    1981-02-01T23:59:59.000Z

    A source book on nuclear dual-purpose electric/distillation desalination plants was prepared to assist government and other planners in preparing broad evaluations of proposed applications of dual-purpose plants. The document is divided into five major sections. Section 1 presents general discussions relating to the benefits of dual-purpose plants, and spectrum for water-to-power ratios. Section 2 presents information on commercial nuclear plants manufactured by US manufacturers. Section 3 gives information on distillation desalting processes and equipment. Section 4 presents a discussion on feedwater pretreatment and scale control. Section 5 deals with methods for coupling the distillation and electrical generating plants to operate in the dual mode.

  2. Use of neural networks in the operation of nuclear power plants

    SciTech Connect (OSTI)

    Uhrig, R.E. (Tennessee Univ., Knoxville, TN (USA) Oak Ridge National Lab., TN (USA))

    1990-01-01T23:59:59.000Z

    Application of neural networks to the operation of nuclear power plants is being investigated under a US Department of Energy sponsored program at the University of Tennessee. Projects include the feasibility of using neural networks for the following tasks: (a) diagnosing specific abnormal conditions, (b) detection of the change of mode of operation, (c) signal validation, (d) monitoring of check valves, (e) modeling of the plant thermodynamics, (f) emulation of core reload calculations, (g) analysis of temporal sequences in NRC's licensee event report,'' (h) monitoring of plant parameters, and (i) analysis of plant vibrations. Each of these projects and its status are described briefly in this article. the objective of each of these projects is to enhance the safety and performance of nuclear plants through the use of neural networks. 6 refs.

  3. Inspection of Nuclear Power Plant Structures - Overview of Methods and Related Applications

    SciTech Connect (OSTI)

    Naus, Dan J [ORNL

    2009-05-01T23:59:59.000Z

    The objectives of this limited study were to provide an overview of the methods that are available for inspection of nuclear power plant reinforced concrete and metallic structures, and to provide an assessment of the status of methods that address inspection of thick, heavily-reinforced concrete and inaccessible areas of the containment metallic pressure boundary. In meeting these objectives a general description of nuclear power plant safety-related structures was provided as well as identification of potential degradation factors, testing and inspection requirements, and operating experience; methods for inspection of nuclear power plant reinforced concrete structures and containment metallic pressure boundaries were identified and described; and applications of nondestructive evaluation methods specifically related to inspection of thick-section reinforced concrete structures and inaccessible portions of containment metallic pressure boundaries were summarized. Recommendations are provided on utilization of test article(s) to further advance nondestructive evaluation methods related to thick-section, heavily-reinforced concrete and inaccessible portions of the metallic pressure boundary representative of nuclear power plant containments. Conduct of a workshop to provide an update on applications and needed developments for nondestructive evaluation of nuclear power plant structures would also be of benefit.

  4. Digital Full-Scope Simulation of a Conventional Nuclear Power Plant Control Room, Phase 2: Installation of a Reconfigurable Simulator to Support Nuclear Plant Sustainability

    SciTech Connect (OSTI)

    Ronald L. Boring; Vivek Agarwal; Kirk Fitzgerald; Jacques Hugo; Bruce Hallbert

    2013-03-01T23:59:59.000Z

    The U.S. Department of Energy’s Light Water Reactor Sustainability program has developed a control room simulator in support of control room modernization at nuclear power plants in the U.S. This report highlights the recent completion of this reconfigurable, full-scale, full-scope control room simulator buildout at the Idaho National Laboratory. The simulator is fully reconfigurable, meaning it supports multiple plant models developed by different simulator vendors. The simulator is full-scale, using glasstop virtual panels to display the analog control boards found at current plants. The present installation features 15 glasstop panels, uniquely achieving a complete control room representation. The simulator is also full-scope, meaning it uses the same plant models used for training simulators at actual plants. Unlike in the plant training simulators, the deployment on glasstop panels allows a high degree of customization of the panels, allowing the simulator to be used for research on the design of new digital control systems for control room modernization. This report includes separate sections discussing the glasstop panels, their layout to mimic control rooms at actual plants, technical details on creating a multi-plant and multi-vendor reconfigurable simulator, and current efforts to support control room modernization at U.S. utilities. The glasstop simulator provides an ideal testbed for prototyping and validating new control room concepts. Equally importantly, it is helping create a standardized and vetted human factors engineering process that can be used across the nuclear industry to ensure control room upgrades maintain and even improve current reliability and safety.

  5. Online Condition Monitoring to Enable Extended Operation of Nuclear Power Plants

    SciTech Connect (OSTI)

    Meyer, Ryan M.; Bond, Leonard J.; Ramuhalli, Pradeep

    2012-03-31T23:59:59.000Z

    Safe, secure, and economic operation of nuclear power plants will remain of strategic significance. New and improved monitoring will likely have increased significance in the post-Fukushima world. Prior to Fukushima, many activities were already underway globally to facilitate operation of nuclear power plants beyond their initial licensing periods. Decisions to shut down a nuclear power plant are mostly driven by economic considerations. Online condition monitoring is a means to improve both the safety and economics of extending the operating lifetimes of nuclear power plants, enabling adoption of proactive aging management. With regard to active components (e.g., pumps, valves, motors, etc.), significant experience in other industries has been leveraged to build the science base to support adoption for online condition-based maintenance and proactive aging management in the nuclear industry. Many of the research needs are associated with enabling proactive management of aging in passive components (e.g., pipes, vessels, cables, containment structures, etc.). This paper provides an overview of online condition monitoring for the nuclear power industry with an emphasis on passive components. Following the overview, several technology/knowledge gaps are identified, which require addressing to facilitate widespread online condition monitoring of passive components.

  6. The Regulatory Challenges of Decommissioning Nuclear Power Plants in Korea - 13101

    SciTech Connect (OSTI)

    Lee, Jungjoon; Ahn, Sangmyeon; Choi, Kyungwoo [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 305-338 (Korea, Republic of)] [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 305-338 (Korea, Republic of); Kim, Juyoul; Kim, Juyub [FNC Technology, 46 Tabsil-ro, Giheung-gu, Yongin 446-902 (Korea, Republic of)] [FNC Technology, 46 Tabsil-ro, Giheung-gu, Yongin 446-902 (Korea, Republic of)

    2013-07-01T23:59:59.000Z

    As of 2012, 23 units of nuclear power plants are in operation, but there is no experience of permanent shutdown and decommissioning of nuclear power plant in Korea. It is realized that, since late 1990's, improvement of the regulatory framework for decommissioning has been emphasized constantly from the point of view of International Atomic Energy Agency (IAEA)'s safety standards. And it is known that now IAEA prepare the safety requirement on decommissioning of facilities, its title is the Safe Decommissioning of Facilities, General Safety Requirement Part 6. According to the result of IAEA's Integrated Regulatory Review Service (IRRS) mission to Korea in 2011, it was recommended that the regulatory framework for decommissioning should require decommissioning plans for nuclear installations to be constructed and operated and these plans should be updated periodically. In addition, after the Fukushima nuclear disaster in Japan in March of 2011, preparedness for early decommissioning caused by an unexpected severe accident became also important issues and concerns. In this respect, it is acknowledged that the regulatory framework for decommissioning of nuclear facilities in Korea need to be improved. First of all, we identify the current status and relevant issues of regulatory framework for decommissioning of nuclear power plants compared to the IAEA's safety standards in order to achieve our goal. And then the plan is to be established for improvement of regulatory framework for decommissioning of nuclear power plants in Korea. After dealing with it, it is expected that the revised regulatory framework for decommissioning could enhance the safety regime on the decommissioning of nuclear power plants in Korea in light of international standards. (authors)

  7. Initiating Event Rates at U.S. Nuclear Power Plants 1988–2013

    SciTech Connect (OSTI)

    John A. Schroeder; Gordon R. Bower

    2014-02-01T23:59:59.000Z

    Analyzing initiating event rates is important because it indicates performance among plants and also provides inputs to several U.S. Nuclear Regulatory Commission (NRC) risk-informed regulatory activities. This report presents an analysis of initiating event frequencies at U.S. commercial nuclear power plants since each plant’s low-power license date. The evaluation is based on the operating experience from fiscal year 1988 through 2013 as reported in licensee event reports. Engineers with nuclear power plant experience staff reviewed each event report since the last update to this report for the presence of valid scrams or reactor trips at power. To be included in the study, an event had to meet all of the following criteria: includes an unplanned reactor trip (not a scheduled reactor trip on the daily operations schedule), sequence of events starts when reactor is critical and at or above the point of adding heat, occurs at a U.S. commercial nuclear power plant (excluding Fort St. Vrain and LaCrosse), and is reported by a licensee event report. This report displays occurrence rates (baseline frequencies) for the categories of initiating events that contribute to the NRC’s Industry Trends Program. Sixteen initiating event groupings are trended and displayed. Initiators are plotted separately for initiating events with different occurrence rates for boiling water reactors and pressurized water reactors. p-values are given for the possible presence of a trend over the most recent 10 years.

  8. Use of fuel cells for improving on-site emergency power availability and reliability ad nuclear power plants

    E-Print Network [OSTI]

    Akkaynak, Derya

    2005-01-01T23:59:59.000Z

    To assure safe shutdown of a nuclear power plant, there must always be reliable means of decay heat removal provided, in last resort, by an Emergency Core Cooling System (ECCS). Currently the majority of nuclear power ...

  9. Volume I, Summary Report: A Roadmap to Deploy New Nuclear Power Plants in the United States by 2010:

    Broader source: Energy.gov [DOE]

    Nuclear power plants in the United States currently produce about 20 percent of the nation’s electricity. This nuclear-generated electricity is safe, clean and economical, and does not emit...

  10. Potential application of neural networks to the operation of nuclear power plants

    SciTech Connect (OSTI)

    Uhrig, R.E. [University of Tennessee, Knoxville, TN (United States)]|[Oak Ridge National Laboratory, TN (United States)

    1991-01-01T23:59:59.000Z

    The application of neural networks, a rapidly evolving technology used extensively in defense applications, to some of the problems of operating nuclear power plants is a logical complement to the expert systems currently being introduced in some of those plants. The potential applications of neural networks include, but are not limited to: (1) Diagnosing specific abnormal conditions. (2) Identifying nonlinear dynamics and transients. (3) Detecting the change of mode of operation. (4) Controlling temperature and pressure during start-up. (5) validating signals. (6) Plant-wide monitoring using autoassociative neural networks. (7) Monitoring of check valves. (8) Modeling the plant thermodynamics to increase efficiency. (9) Emulating core reload calculations. (10) Analyzing temporal sequences in the U.S. Nuclear Regulatory Commission Licensee Event Reports. (11) Monitoring plant parameters. (12) Analyzing vibrations in plants and rotating machinery. The work on such applications indicates that neural networks alone, or in conjunction with other advanced technologies, have the potential to enhance the safety, reliability, and operability of nuclear power plants. 36 refs.

  11. OVERVIEW OF A RECONFIGURABLE SIMULATOR FOR MAIN CONTROL ROOM UPGRADES IN NUCLEAR POWER PLANTS

    SciTech Connect (OSTI)

    Ronald L. Boring

    2012-10-01T23:59:59.000Z

    This paper provides background on a reconfigurable control room simulator for nuclear power plants. The main control rooms in current nuclear power plants feature analog technology that is growing obsolete. The need to upgrade control rooms serves the practical need of maintainability as well as the opportunity to implement newer digital technologies with added functionality. There currently exists no dedicated research simulator for use in human factors design and evaluation activities for nuclear power plant modernization in the U.S. The new research simulator discussed in this paper provides a test bed in which operator performance on new control room concepts can be benchmarked against existing control rooms and in which new technologies can be validated for safety and usability prior to deployment.

  12. Safeguards Guidance Document for Designers of Commercial Nuclear Facilities: International Nuclear Safeguards Requirements and Practices For Uranium Enrichment Plants

    SciTech Connect (OSTI)

    Robert Bean; Casey Durst

    2009-10-01T23:59:59.000Z

    This report is the second in a series of guidelines on international safeguards requirements and practices, prepared expressly for the designers of nuclear facilities. The first document in this series is the description of generic international nuclear safeguards requirements pertaining to all types of facilities. These requirements should be understood and considered at the earliest stages of facility design as part of a new process called “Safeguards-by-Design.” This will help eliminate the costly retrofit of facilities that has occurred in the past to accommodate nuclear safeguards verification activities. The following summarizes the requirements for international nuclear safeguards implementation at enrichment plants, prepared under the Safeguards by Design project, and funded by the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA), Office of NA-243. The purpose of this is to provide designers of nuclear facilities around the world with a simplified set of design requirements and the most common practices for meeting them. The foundation for these requirements is the international safeguards agreement between the country and the International Atomic Energy Agency (IAEA), pursuant to the Treaty on the Non-proliferation of Nuclear Weapons (NPT). Relevant safeguards requirements are also cited from the Safeguards Criteria for inspecting enrichment plants, found in the IAEA Safeguards Manual, Part SMC-8. IAEA definitions and terms are based on the IAEA Safeguards Glossary, published in 2002. The most current specification for safeguards measurement accuracy is found in the IAEA document STR-327, “International Target Values 2000 for Measurement Uncertainties in Safeguarding Nuclear Materials,” published in 2001. For this guide to be easier for the designer to use, the requirements have been restated in plainer language per expert interpretation using the source documents noted. The safeguards agreement is fundamentally a legal document. As such, it is written in a legalese that is understood by specialists in international law and treaties, but not by most outside of this field, including designers of nuclear facilities. For this reason, many of the requirements have been simplified and restated. However, in all cases, the relevant source document and passage is noted so that readers may trace the requirement to the source. This is a helpful living guide, since some of these requirements are subject to revision over time. More importantly, the practices by which the requirements are met are continuously modernized by the IAEA and nuclear facility operators to improve not only the effectiveness of international nuclear safeguards, but also the efficiency. As these improvements are made, the following guidelines should be updated and revised accordingly.

  13. The status of nuclear power plants in the People's Republic of China

    SciTech Connect (OSTI)

    Puckett, J.

    1991-05-01T23:59:59.000Z

    China's main energy source is coal, but transportation and environmental problems make that fuel less than desirable. Therefore, the Chinese, as part of an effort toward alternative energy sources, are developing nuclear power plants. In addition to providing a cleaner power source, development of nuclear energy would improve the Chinese economic condition and give the nation greater world status. China's first plants, at Qinshan and Daya Bay, are still incomplete. However, China is working toward completion of those reactors and planning the training and operating procedures needed to operate them. At the same time, it is improving its nuclear fuel exports. As they develop the capability for generating nuclear power, the Chinese seem to be aware of the accompanying quality and safety considerations, which they have declared to be first priorities. 50 refs., 7 figs.

  14. Nuclear safety procedure upgrade project at USEC/MMUS gaseous diffusion plants

    SciTech Connect (OSTI)

    Kocsis, F.J. III

    1994-12-31T23:59:59.000Z

    Martin Marietta Utility Services has embarked on a program to upgrade procedures at both of its Gaseous Diffusion Plant sites. The transition from a U.S. Department of Energy government-operated facility to U.S. Nuclear Regulatory Commission (NRC) regulated has necessitated a complete upgrade of plant operating procedures and practices incorporating human factors as well as a philosophy change in their use. This program is designed to meet the requirements of the newly written 10CFR76, {open_quotes}The Certification of Gaseous Diffusion Plants,{close_quotes} and aid in progression toward NRC certification. A procedures upgrade will help ensure increased nuclear safety, enhance plant operation, and eliminate personnel procedure errors/occurrences.

  15. Aging management guideline for commercial nuclear power plants - tanks and pools

    SciTech Connect (OSTI)

    Blocker, E.; Smith, S.; Philpot, L.; Conley, J.

    1996-02-01T23:59:59.000Z

    Continued operation of nuclear power plants for periods that extend beyond their original 40-year license period is a desirable option for many U.S. utilities. U.S. Nuclear Regulatory Commission (NRC) approval of operating license renewals is necessary before continued operation becomes a reality. Effective aging management for plant components is important to reliability and safety, regardless of current plant age or extended life expectations. However, the NRC requires that aging evaluations be performed and the effectiveness of aging management programs be demonstrated for components considered within the scope of license renewal before granting approval for operation beyond 40 years. Both the NRC and the utility want assurance that plant components will be highly reliable during both the current license term and throughout the extended operating period. In addition, effective aging management must be demonstrated to support Maintenance Rule (10 CFR 50.65) activities.

  16. Improved assessment of population doses and risk factors for a nuclear power plant under accident conditions

    E-Print Network [OSTI]

    Meyer, Christopher Martin

    1985-01-01T23:59:59.000Z

    In order to assess the doses received by the members of the public due to an accident at a nuclear power plant, a number of physical processes must be modeled. These processes include the release of radioactive materials, the atmospheric dispersion... representative of the industry. Generic reactor sites must be conceptualized in order to obtain meteorologic data which is representative of the areas within the United States in which nuclear power facilities have been sited, Information such as population...

  17. Nuclear criticality safety evaluation of Spray Booth Operations in X-705, Portsmouth Gaseous Diffusion Plant

    SciTech Connect (OSTI)

    Sheaffer, M.K.; Keeton, S.C.

    1993-09-20T23:59:59.000Z

    This report evaluates nuclear criticality safety for Spray Booth Operations in the Decontamination and Recovery Facility, X-705, at the Portsmouth Gaseous Diffusion Plant. A general description of current procedures and related hardware/equipment is presented. Control parameters relevant to nuclear criticality safety are explained, and a consolidated listing of administrative controls and safety systems is developed. Based on compliance with DOE Orders and MMES practices, the overall operation is evaluated, and recommendations for enhanced safety are suggested.

  18. An analysis of nuclear power plant operating costs: A 1995 update

    SciTech Connect (OSTI)

    NONE

    1995-04-21T23:59:59.000Z

    Over the years real (inflation-adjusted) O&M cost have begun to level off. The objective of this report is to determine whether the industry and NRC initiatives to control costs have resulted in this moderation in the growth of O&M costs. Because the industry agrees that the control of O&M costs is crucial to the viability of the technology, an examination of the factors causing the moderation in costs is important. A related issue deals with projecting nuclear operating costs into the future. Because of the escalation in nuclear operating costs (and the fall in fossil fuel prices) many State and Federal regulatory commissions are examining the economics of the continued operation of nuclear power plants under their jurisdiction. The economics of the continued operation of a nuclear power plant is typically examined by comparing the cost of the plants continued operation with the cost of obtaining the power from other sources. This assessment requires plant-specific projections of nuclear operating costs. Analysts preparing these projections look at past industry-wide cost trends and consider whether these trends are likely to continue. To determine whether these changes in trends will continue into the future, information about the causal factors influencing costs and the future trends in these factors are needed. An analysis of the factors explaining the moderation in cost growth will also yield important insights into the question of whether these trends will continue.

  19. Community emergency response to nuclear power plant accidents: A selected and partially annotated bibliography

    SciTech Connect (OSTI)

    Youngen, G.

    1988-10-01T23:59:59.000Z

    The role of responding to emergencies at nuclear power plants is often considered the responsibility of the personnel onsite. This is true for most, if not all, of the incidents that may happen during the course of the plant`s operating lifetime. There is however, the possibility of a major accident occurring at anytime. Major nuclear accidents at Chernobyl and Three Mile Island have taught their respective countries and communities a significant lesson in local emergency preparedness and response. Through these accidents, the rest of the world can also learn a great deal about planning, preparing and responding to the emergencies unique to nuclear power. This bibliography contains books, journal articles, conference papers and government reports on emergency response to nuclear power plant accidents. It does not contain citations for ``onsite`` response or planning, nor does it cover the areas of radiation releases from transportation accidents. The compiler has attempted to bring together a sampling of the world`s collective written experience on dealing with nuclear reactor accidents on the sate, local and community levels. Since the accidents at Three Mile Island and Chernobyl, that written experience has grown enormously.

  20. Natural Circulation in Water Cooled Nuclear Power Plants Phenomena, models, and methodology for system reliability assessments

    SciTech Connect (OSTI)

    Jose Reyes

    2005-02-14T23:59:59.000Z

    In recent years it has been recognized that the application of passive safety systems (i.e., those whose operation takes advantage of natural forces such as convection and gravity), can contribute to simplification and potentially to improved economics of new nuclear power plant designs. In 1991 the IAEA Conference on ''The Safety of Nuclear Power: Strategy for the Future'' noted that for new plants the use of passive safety features is a desirable method of achieving simplification and increasing the reliability of the performance of essential safety functions, and should be used wherever appropriate''.

  1. Guidance for Deployment of Mobile Technologies for Nuclear Power Plant Field Workers

    SciTech Connect (OSTI)

    Heather D. Medema; Ronald K. Farris

    2012-09-01T23:59:59.000Z

    This report is a guidance document prepared for the benefit of commercial nuclear power plants’ (NPPs) supporting organizations and personnel who are considering or undertaking deployment of mobile technology for the purpose of improving human performance and plant status control (PSC) for field workers in an NPP setting. This document especially is directed at NPP business managers, Electric Power Research Institute, Institute of Nuclear Power Operations, and other non-Information Technology personnel. This information is not intended to replace basic project management practices or reiterate these processes, but is to support decision-making, planning, and preparation of a business case.

  2. Transgenic plants are sensitive bioindicators of nuclear pollution caused by the Chernobyl accident

    SciTech Connect (OSTI)

    Kovalchuk, I.; Kovalchuk, O. [Ivano-Frankivsk State Medical Academy (Ukraine)]|[Friedrich Miescher Inst., Basel (Switzerland); Arkhipov, A. [Chernobyl Scientific and Technical Center of International Research (Ukraine); Hohn, B. [Friedrich Miescher Inst., Basel (Switzerland)

    1998-11-01T23:59:59.000Z

    To evaluate the genetic consequences of radioactive contamination originating from the Nuclear reactor accident of Chernobyl on indigenous populations of plants and animals, it is essential to determine the rates of accumulating genetic changes in chronically irradiated populations. An increase in germline mutation rates in humans living close to the Chernobyl Nuclear Power Plant site, and a two- to tenfold increase in germline mutations in barn swallows breeding in Chernobyl have been reported. Little is known, however, about the effects of chronic irradiation on plant genomes. Ionizing radiation causes double-strand breaks in DNA, which are repaired via illegitimate or homologous recombination. The authors make use of Arabidopsis thaliana plants carrying a {beta}-glucuronidase marker gene as a recombination substrate to monitor genetic alterations in plant populations, which are caused by nuclear pollution of the environment around Chernobyl. A significant increase in somatic intrachromosomal recombination frequencies was observed at nuclear pollution levels from 0.1--900 Ci/km{sup 2}, consistent with an increase in chromosomal aberrations. This bioindicator may serve as a convenient and ethically acceptable alternative to animal systems.

  3. DOE Makes Available $8 Million for Pre-Conceptual Design of Next...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    for Proposals for Engineering and Design on NGNP DOE Seeks Industry Participation for Engineering Services to Design Next Generation Nuclear Plant Industry Participation Sought...

  4. Microsoft Word - 911144_0_SSC-13 S-I HPS Test Plan_rel.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    4 Revision 0 ENGINEERING SERVICES FOR THE NEXT GENERATION NUCLEAR PLANT (NGNP) WITH HYDROGEN PRODUCTION Test Plan for S-I Hydrogen Production System (HPS) Prepared by General...

  5. Microsoft Word - PC-000587_0_rel.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    7 Revision 0 Engineering Services for the Next Generation Nuclear Plant (NGNP) with Hydrogen Production Assessment of Russian Federation Test Facilities Capabilities to Support...

  6. Microsoft Word - 911136_0 SSC-4b Reactor Graphite Test Plan_rel...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    6 Revision 0 Engineering Services for the Next Generation Nuclear Plant (NGNP) with Hydrogen Production Test Plan for Reactor Graphite Elements Prepared by General Atomics for the...

  7. Microsoft Word - 911141_0_SSC-9 RCCS Test Plan.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    1 Revision 0 ENGINEERING SERVICES FOR THE NEXT GENERATION NUCLEAR PLANT (NGNP) WITH HYDROGEN PRODUCTION Test Plan for the Reactor Cavity Cooling System Prepared by General Atomics...

  8. Microsoft Word - 911145_0_SSC-14 FHSS Test Plan_rel.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    5 Revision 0 ENGINEERING SERVICES FOR THE NEXT GENERATION NUCLEAR PLANT (NGNP) WITH HYDROGEN PRODUCTION Test Plan - Fuel Handling and Storage System Prepared by General Atomics For...

  9. Microsoft Word - 911138_0_SSC-6 Helium Circulator Test Plan_rel...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    8 Revision 0 ENGINEERING SERVICES FOR THE NEXT GENERATION NUCLEAR PLANT (NGNP) WITH HYDROGEN PRODUCTION Test Plan for Helium Circulators (PHTS, SCS, SHTS) Prepared by General...

  10. Microsoft Word - 911135_0 SSC-4a Reactor Core Test Plan_rel.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    5 Revision 0 ENGINEERING SERVICES FOR THE NEXT GENERATION NUCLEAR PLANT (NGNP) WITH HYDROGEN PRODUCTION Test Plan for the Reactor Core Assembly Prepared by General Atomics For the...

  11. Microsoft Word - 911140_0_SCHE Test Plan_rel.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    0 Revision 0 ENGINEERING SERVICES FOR THE NEXT GENERATION NUCLEAR PLANT (NGNP) WITH HYDROGEN PRODUCTION Test Plan - Shutdown Cooling Heat Exchanger Prepared by General Atomics For...

  12. Microsoft Word - 911133_0 SSC-1 Test Plan_rel.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    3 Revision 0 ENGINEERING SERVICES FOR THE NEXT GENERATION NUCLEAR PLANT (NGNP) WITH HYDROGEN PRODUCTION Test Plan for Reactor Control Equipment Prepared by General Atomics For the...

  13. Microsoft Word - 911139_0_SSC-7 IHX Test Plan_rel.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    9 Revision 0 ENGINEERING SERVICES FOR THE NEXT GENERATION NUCLEAR PLANT (NGNP) WITH HYDROGEN PRODUCTION Test Plan - Intermediate Heat Exchanger PCHE Type Prepared by General...

  14. Microsoft Word - 911134_0 SSC-2 Control Rod Test Plan_rel.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    4 Revision 0 ENGINEERING SERVICES FOR THE NEXT GENERATION NUCLEAR PLANT (NGNP) WITH HYDROGEN PRODUCTION Test Plan - Control Rods Prepared by General Atomics For the Battelle Energy...

  15. Microsoft Word - 911143_0_SSC-11 PCS Test Plan_rel.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    3 Revision 0 ENGINEERING SERVICES FOR THE NEXT GENERATION NUCLEAR PLANT (NGNP) WITH HYDROGEN PRODUCTION Test Plan - Power Conversion System Equipment for a Direct Combined Cycle...

  16. Microsoft Word - 911137_0 SSC-5 RPV Test Plan_rel.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    7 Revision 0 ENGINEERING SERVICES FOR THE NEXT GENERATION NUCLEAR PLANT (NGNP) WITH HYDROGEN PRODUCTION Test Plan - Reactor Pressure Vessel Prepared by General Atomics For the...

  17. Microsoft Word - 911147_0_SSC-16 RPS IPS and PCDIS Test Plan...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    7 Revision 0 ENGINEERING SERVICES FOR THE NEXT GENERATION NUCLEAR PLANT (NGNP) WITH HYDROGEN PRODUCTION Test Plan for RPS, IPS and PCDIS Prepared by General Atomics For the...

  18. Microsoft Word - INL-ext-09-16778.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    ix ACRONYMS CD compact disk DDN Design Data Need NGNP Next Generation Nuclear Plant PIRT Phenomena Identification and Ranking Table R&D research and development RMS Risk...

  19. Microsoft Word - PLN-4086 Rev 1.docx

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Test Reactor IAEA International Atomic Energy Agency MHTGR modular high temperature gas-cooled reactor NGNP Next Generation Nuclear Plant PSID preliminary safety information...

  20. Microsoft Word - TEV 704 Rev 2.docx

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Generation Nuclear Plant (NGNP) Project to evaluate the integration of high-temperature gas-cooled reactor (HTGR) technology with conventional chemical processes. This TEV...

  1. Managing aging in nuclear power plants: Insights from NRC maintenance team inspection reports

    SciTech Connect (OSTI)

    Fresco, A.; Subudhi, M.; Gunther, W.; Grove, E.; Taylor, J. [Brookhaven National Lab., Upton, NY (United States)

    1993-12-01T23:59:59.000Z

    A plant`s maintenance program is the principal vehicle through which age-related degradation is managed. From 1988 to 1991, the NRC evaluated the maintenance program of every nuclear power plant in the United States. Forty-four out of a total of 67 of the reports issued on these in-depth team inspections were reviewed for insights into the strengths and weaknesses of the programs as related to the need to understand and manage the effects of aging on nuclear plant systems, structures, and components. Relevant information was extracted from these inspection reports and sorted into several categories, including Specific Aging Insights, Preventive Maintenance, Predictive Maintenance and Condition Monitoring, Post Maintenance Testing, Failure Trending, Root Cause Analysis and Usage of Probabilistic Risk Assessment in the Maintenance Process. Specific examples of inspection and monitoring techniques successfully used by utilities to detect degradation due to aging have been identified. The information also was sorted according to systems and components, including: Auxiliary Feedwater, Main Feedwater, High Pressure Injection for both BWRs and PWRs, Service Water, Instrument Air, and Emergency Diesel Generator Air Start Systems, and Emergency Diesel Generators Air Start Systems, emergency diesel generators, electrical components such as switchgear, breakers, relays, and motor control centers, motor operated valves and check valves. This information was compared to insights gained from the Nuclear Plant Aging Research (NPAR) Program. Attributes of plant maintenance programs where the NRC inspectors felt that improvement was needed to properly address the aging issue also are discussed.

  2. COOLING WATER ISSUES AND OPPORTUNITIES AT U.S. NUCLEAR POWER PLANTS

    SciTech Connect (OSTI)

    Gary Vine

    2010-12-01T23:59:59.000Z

    This report has been prepared for the Department of Energy, Office of Nuclear Energy (DOE-NE), for the purpose of providing a status report on the challenges and opportunities facing the U.S. commercial nuclear energy industry in the area of plant cooling water supply. The report was prompted in part by recent Second Circuit and Supreme Court decisions regarding cooling water system designs at existing thermo-electric power generating facilities in the U.S. (primarily fossil and nuclear plants). At issue in the courts have been Environmental Protection Agency regulations that define what constitutes “Best Technology Available” for intake structures that withdraw cooling water that is used to transfer and reject heat from the plant’s steam turbine via cooling water systems, while minimizing environmental impacts on aquatic life in nearby water bodies used to supply that cooling water. The report was also prompted by a growing recognition that cooling water availability and societal use conflicts are emerging as strategic energy and environmental issues, and that research and development (R&D) solutions to emerging water shortage issues are needed. In particular, cooling water availability is an important consideration in siting decisions for new nuclear power plants, and is an under-acknowledged issue in evaluating the pros and cons of retrofitting cooling towers at existing nuclear plants. Because of the significant ongoing research on water issues already being performed by industry, the national laboratories and other entities, this report relies heavily on ongoing work. In particular, this report has relied on collaboration with the Electric Power Research Institute (EPRI), including its recent work in the area of EPA regulations governing intake structures in thermoelectric cooling water systems.

  3. EDF Nuclear Power Plants Operating Experience with MOX fuel

    SciTech Connect (OSTI)

    Thibault, Xavier [EDF Generation, Tour EDF Part Dieu - 9 rue des Cuirassiers B.P.3181 - 69402 Lyon Cedex 03 (France)

    2006-07-01T23:59:59.000Z

    EDF started Plutonium recycling in PWR in 1987 and progressively all the 20 reactors, licensed in using MOX fuel, have been loaded with MOX assemblies. At the origin of MOX introduction, these plants operated at full power in base load and the core management limited the irradiation time of MOX fuel assemblies to 3 annual cycles. Since 1995 all these reactors can operate in load follow mode. Since that time, a large amount of experience has been accumulated. This experience is very positive considering: - Receipt, handling, in core behaviour, pool storage and shipment of MOX fuel; - Operation of the various systems of the plant; - Environment impact; - Radioprotection; - Safety file requirements; - Availability for the grid. In order to reduce the fuel cost and to reach a better adequacy between UO{sub 2} fuel reprocessing flow and plutonium consumption, EDF had decided to improve the core management of MOX plants. This new core management call 'MOX Parity' achieves parity for MOX and UO{sub 2} assemblies in term of discharge burn-up. Compared to the current MOX assembly the Plutonium content is increased from 7,08% to 8,65% (equivalent to natural uranium enriched to respectively 3,25% and 3,7%) and the maximum MOX assembly burn-up moves from 42 to 52 GWd/t. This amount of burn-up is obtained from loading MOX assemblies for one additional annual cycle. Some, but limited, adaptations of the plant are necessary. In addition a new MOX fuel assembly has been designed to comply with the safety criteria taking into account the core management performances. These design improvements are based on the results of an important R and D program including numerous experimental tests and post-irradiated fuel examinations. In particular, envelope conditions compared to MOX Parity neutronic solicitations has been extensively investigated in order to get a full knowledge of the in reactor fuel behavior. Moreover, the operating conditions of the plant have been evaluated in many details and finally no important impact is anticipated. The industrial maturity of plutonium recycling activities is fully demonstrated and a new progress can be done with a complete confidence. The licensing process of 'MOX Parity' core management is in progress and its implementation on the 20 PWR is now expected at mid 2007. (author)

  4. Physical protection solutions for security problems at nuclear power plants. [PWR; BWR

    SciTech Connect (OSTI)

    Darby, J.L.; Jacobs, J.

    1980-09-01T23:59:59.000Z

    Under Department of Energy sponsorship, Sandia National Laboratories has developed a broad technological base of components and integrated systems to address security concerns at facilities of importance, including nuclear reactors. The primary security concern at a light water reactor is radiological sabotage, a deliberate set of actions at a plant which could expose the public to a significant amount of radiation (on the order of 10 CFR 100 limits). (Also of importance to plant operators are acts of industrial sabotage that could prevent a plant from producing electrical power).

  5. The integrated workstation: A common, consistent link between nuclear plant personnel and plant information and computerized resources

    SciTech Connect (OSTI)

    Wood, R.T.; Knee, H.E.; Mullens, J.A.; Munro, J.K. Jr.; Swail, B.K.; Tapp, P.A.

    1993-05-01T23:59:59.000Z

    The increasing use of computer technology in the US nuclear power industry has greatly expanded the capability to obtain, analyze, and present data about the plant to station personnel. Data concerning a power plant`s design, configuration, operational and maintenance histories, and current status, and the information that can be derived from them, provide the link between the plant and plant staff. It is through this information bridge that operations, maintenance and engineering personnel understand and manage plant performance. However, it is necessary to transform the vast quantity of data available from various computer systems and across communications networks into clear, concise, and coherent information. In addition, it is important to organize this information into a consolidated, structured form within an integrated environment so that various users throughout the plant have ready access at their local station to knowledge necessary for their tasks. Thus, integrated workstations are needed to provide the inquired information and proper software tools, in a manner that can be easily understood and used, to the proper users throughout the plant. An effort is underway at the Oak Ridge National Laboratory to address this need by developing Integrated Workstation functional requirements and implementing a limited-scale prototype demonstration. The integrated Workstation requirements will define a flexible, expandable computer environment that permits a tailored implementation of workstation capabilities and facilitates future upgrades to add enhanced applications. The functionality to be supported by the integrated workstation and inherent capabilities to be provided by the workstation environment win be described. In addition, general technology areas which are to be addressed in the Integrated Workstation functional requirements will be discussed.

  6. EU could go it alone on nuclear fusion plant 29.11.2004 -10:02 CET | By Richard Carter

    E-Print Network [OSTI]

    EU could go it alone on nuclear fusion plant 29.11.2004 - 10:02 CET | By Richard Carter The EU if no agreement can be reached with Japan by the end of the year over where to build the plant, according to EU research ministers. Talks over the world's first nuclear fusion reactor have stalled because Japan

  7. Ice Thermal Storage Systems for Nuclear Power Plant Supplemental Cooling and Peak Power Shifting

    SciTech Connect (OSTI)

    Haihua Zhao; Hongbin Zhang; Phil Sharpe; Blaise Hamanaka; Wei Yan; WoonSeong Jeong

    2013-03-01T23:59:59.000Z

    Availability of cooling water has been one of the major issues for the nuclear power plant site selection. Cooling water issues have frequently disrupted the normal operation at some nuclear power plants during heat waves and long draught. One potential solution is to use ice thermal storage (ITS) systems that reduce cooling water requirements and boost the plant’s thermal efficiency in hot hours. ITS uses cheap off-peak electricity to make ice and uses the ice for supplemental cooling during peak demand time. ITS also provides a way to shift a large amount of electricity from off peak time to peak time. For once-through cooling plants near a limited water body, adding ITS can bring significant economic benefits and avoid forced derating and shutdown during extremely hot weather. For the new plants using dry cooling towers, adding the ITS systems can effectively reduce the efficiency loss during hot weather so that new plants could be considered in regions lack of cooling water. This paper will review light water reactor cooling issues and present the feasibility study results.

  8. U.S. Nuclear Power Plant Operating Cost and Experience Summaries

    SciTech Connect (OSTI)

    Reid, RL

    2003-09-18T23:59:59.000Z

    The ''U.S. Nuclear Power Plant Operating Cost and Experience Summaries'' (NUREG/CR-6577, Supp. 2) report has been prepared to provide historical operating cost and experience information on U.S. commercial nuclear power plants during 2000-2001. Costs incurred after initial construction are characterized as annual production costs, which represent fuel and plant operating and maintenance expenses, and capital expenditures related to facility additions/modifications, which are included in the plant capital asset base. As discussed in the report, annual data for these two cost categories were obtained from publicly available reports and must be accepted as having different degrees of accuracy and completeness. Treatment of inconclusive and incomplete data is discussed. As an aid to understanding the fluctuations in the cost histories, operations summaries for each nuclear unit are provided. The intent of these summaries is to identify important operating events; refueling, major maintenance, and other significant outages; operating milestones; and significant licensing or enforcement actions. Information used in the summaries is condensed from operating reports submitted by the licensees, the Nuclear Regulatory Commission (NRC) database for enforcement actions, and outage reports.

  9. Aerial Radiation Measurements from the Fukushima Dai-ichi Nuclear Power Plant Accident

    SciTech Connect (OSTI)

    Guss, P. P.

    2012-07-16T23:59:59.000Z

    This document is a slide show type presentation concerning DOE and Aerial Measuring System (AMS) activities and results with respect to assessing the consequences of the releases from the Fukushima Dai-ichi Nuclear Power Plant. These include ground monitoring and aerial monitoring.

  10. Development of Advanced Technologies to Reduce Design, Fabrication and Construction Costs for Future Nuclear Power Plants

    SciTech Connect (OSTI)

    Camillo A. DiNunzio Framatome ANP DE& S; Dr. Abhinav Gupta Assistant Professor NCSU; Dr. Michael Golay Professor MIT Dr. Vincent Luk Sandia National Laboratories; Rich Turk Westinghouse Electric Company Nuclear Systems; Charles Morrow, Sandia National Laboratories; Geum-Taek Jin, Korea Power Engineering Company Inc.

    2002-11-30T23:59:59.000Z

    OAK-B135 This report presents a summation of the third and final year of a three-year investigation into methods and technologies for substantially reducing the capital costs and total schedule for future nuclear plants. In addition, this is the final technical report for the three-year period of studies.

  11. Survey of thermal-hydraulic models of commercial nuclear power plants

    SciTech Connect (OSTI)

    Determan, J.C.; Hendrix, C.E.

    1992-12-01T23:59:59.000Z

    A survey of the thermal-hydraulic models of nuclear power plants has been performed to identify the NRC's current analytical capabilities for critical event response. The survey also supports ongoing research for accident management. The results of the survey are presented here. The PC database which records detailed data on each model is described.

  12. Survey of thermal-hydraulic models of commercial nuclear power plants

    SciTech Connect (OSTI)

    Determan, J.C.; Hendrix, C.E.

    1992-12-01T23:59:59.000Z

    A survey of the thermal-hydraulic models of nuclear power plants has been performed to identify the NRC`s current analytical capabilities for critical event response. The survey also supports ongoing research for accident management. The results of the survey are presented here. The PC database which records detailed data on each model is described.

  13. Nuclear Plant Feedwater Heater Handbook. Volume 2. Design and procurement guidelines. Final report

    SciTech Connect (OSTI)

    Bell, R.J.; Wells, T.G. Jr.

    1985-06-01T23:59:59.000Z

    This document is the second part of a three volume handbook covering closed feedwater heaters for nuclear electric power generating plants. This second volume covers the engineering, design, specification, bid, procurement, fabrication, installation and commissioning of the heater. 22 refs., 12 figs., 17 tabs.

  14. Ranking of four potential nuclear power plant sites in Iraq according to the collective dose criterion

    SciTech Connect (OSTI)

    Marouf, B.A.; Al-Kateeb, G.H.; Al-Ani, D.S. [and others

    1991-07-01T23:59:59.000Z

    The collective dose criterion was used to rank four potential nuclear power-plant sites. Baiji, Al-Mahzam, Al-Abbasia, and Abu-Dalaf. Atmospheric as well as aquatic releases of radionuclides into the environment from the VVER 440 nuclear power plant during normal operation were used to estimate the collective dose equivalents. The results indicated that the collective doses at Baiji, Al-Mahzam, Al-Abbasia, and Abu-Dalaf were 3.6 x 10{sup -2}, 4.7 x 10{sup -2}, 1.1 x 10{sup -1}, and 1.2 x 10{sup -1} man-Sv, respectively. Thus the order of preference is Baiji, Al-Mahzam, Al-Abbasia, and Abu-Dalaf. The effective dose equivalents to the highest exposed individual resulting from atmospheric as well as aquatic releases of radionuclides from the reactor at any one of the four potential nuclear power-plant sites would not exceed 2 x 10{sup -5} Sv/yr. Thus any one of the four sites is suitable for the operation of the 440 nuclear power plants. 27 refs., 1 tab.

  15. Aging management of nuclear power plant containments for license renewal

    SciTech Connect (OSTI)

    Liu, W.C.; Kuo, P.T.; Lee, S.S.

    1997-09-01T23:59:59.000Z

    In 1990, the Nuclear Management and Resources Council (NUMARC), now the Nuclear Energy Institute (NEI), submitted for NRC review, the industry reports (IRs), NUMARC Report 90-01 and NUMARC Report 90-10, addressing aging management issues associated with PWR containments and BWR containments for license renewal, respectively. In 1996, the Commission amended 10 CFR 50.55a to promulgate requirements for inservice inspection of containment structures. This rule amendment incorporates by reference the 1992 Edition with the 1992 Addenda of Subsections IWE and IWL of the ASME Code addressing the inservice inspection of metal containments/liners and concrete containments, respectively. The purpose of this report is to reconcile the technical information and agreements resulting from the NUMARC IR reviews which are generally described in NUREG-1557 and the inservice inspection requirements of subsections IWE and IWL as promulgated in {section}50.55a for license renewal consideration. This report concludes that Subsections IWE and IWL as endorsed in {section}50.55a are generally consistent with the technical agreements reached during the IR reviews. Specific exceptions are identified and additional evaluations and augmented inspections for renewal are recommended.

  16. Prognostics and Health Management in Nuclear Power Plants: A Review of Technologies and Applications

    SciTech Connect (OSTI)

    Coble, Jamie B.; Ramuhalli, Pradeep; Bond, Leonard J.; Hines, Wes; Upadhyaya, Belle

    2012-07-17T23:59:59.000Z

    This report reviews the current state of the art of prognostics and health management (PHM) for nuclear power systems and related technology currently applied in field or under development in other technological application areas, as well as key research needs and technical gaps for increased use of PHM in nuclear power systems. The historical approach to monitoring and maintenance in nuclear power plants (NPPs), including the Maintenance Rule for active components and Aging Management Plans for passive components, are reviewed. An outline is given for the technical and economic challenges that make PHM attractive for both legacy plants through Light Water Reactor Sustainability (LWRS) and new plant designs. There is a general introduction to PHM systems for monitoring, fault detection and diagnostics, and prognostics in other, non-nuclear fields. The state of the art for health monitoring in nuclear power systems is reviewed. A discussion of related technologies that support the application of PHM systems in NPPs, including digital instrumentation and control systems, wired and wireless sensor technology, and PHM software architectures is provided. Appropriate codes and standards for PHM are discussed, along with a description of the ongoing work in developing additional necessary standards. Finally, an outline of key research needs and opportunities that must be addressed in order to support the application of PHM in legacy and new NPPs is presented.

  17. Knowledge and abilities catalog for nuclear power plant operators: boiling water reactors

    SciTech Connect (OSTI)

    Not Available

    1986-09-01T23:59:59.000Z

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWR) (NUREG-1123) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog and Examiners' Handbook for Developing Operator Licensing Examinations (NUREG-1121) will cover those topics listed under Title 10, Code of Federal Regulations, Part 55. The BWR Catalog contains approximately 7000 knowledge and ability (K/A) statements for ROs and SROs at boiling water reactors. Each K/A statement has been rated for its importance to the safe operation of the plant in a manner ensuring personnel and public health and safety. The BWR K/A Catalog is organized into five major sections: Plant-wide Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Function, Emergency and Abnormal Plant Evolutions, Components, and Theory. The BWR Catalog represents a modification of the form and content of the K/A Catalog for Nuclear Power Plant Operators: Pressurized Water Reactors (NUREG-1122). First, categories of knowledge and ability statements have been redefined. Second, the scope of the definition of emergency and abnormal plant evolutions has been revised in line with a symptom-based approach. Third, K/As related to the operational applications of theory have been incorporated into the delineations for both plant systems and emergency and abnormal plant evolutions, while K/As pertaining to theory fundamental to plant operation have been delineated in a separate theory section. Finally, the components section has been revised.

  18. Assessment of inservice conditions of safety-related nuclear plant structures

    SciTech Connect (OSTI)

    Ashar, H.; Bagchi, G.

    1995-06-01T23:59:59.000Z

    The report is a compilation from a number of sources of information related to the condition Of structures and civil engineering features at operating nuclear power plants in the United States. The most significant information came from the hands-on inspection of the six old plants (licensed prior to 1977) performed by the staff of the Civil Engineering and Geosciences Branch (ECGB) in the Division of Engineering of the Office of Nuclear Reactor Regulation. For the containment structures, most of the information related to the degraded conditions came from the licensees as part of the Licensing Event Report System (10 CFR 50.73), or as part of the requirement under limiting condition of operation of the plant-specific Technical Specifications. Most of the information related to the degradation of other Structures and civil engineering features was extracted from the industry survey, the reported incidents, and the plant visits. The report discusses the condition of the structures and civil engineering features at operating nuclear power plants and provides information that would help detect, alleviate, and correct the degraded conditions of the structures and civil engineering features.

  19. Application of NUREG/CR-5999 interim fatigue curves to selected nuclear power plant components

    SciTech Connect (OSTI)

    Ware, A.G.; Morton, D.K.; Nitzel, M.E. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

    1995-03-01T23:59:59.000Z

    Recent test data indicate that the effects of the light water reactor (LWR) environment could significantly reduce the fatigue resistance of materials used in the reactor coolant pressure boundary components of operating nuclear power plants. Argonne National Laboratory has developed interim fatigue curves based on test data simulating LWR conditions, and published them in NUREG/CR-5999. In order to assess the significance of these interim fatigue curves, fatigue evaluations of a sample of the components in the reactor coolant pressure boundary of LWRs were performed. The sample consists of components from facilities designed by each of the four U.S. nuclear steam supply system vendors. For each facility, six locations were studied, including two locations on the reactor pressure vessel. In addition, there are older vintage plants where components of the reactor coolant pressure boundary were designed to codes that did not require an explicit fatigue analysis of the components. In order to assess the fatigue resistance of the older vintage plants, an evaluation was also conducted on selected components of three of these plants. This report discusses the insights gained from the application of the interim fatigue curves to components of seven operating nuclear power plants.

  20. Design issues concerning Iran`s Bushehr nuclear power plant VVER-1000 conversion

    SciTech Connect (OSTI)

    Carson, C.F. [Lawrence Livermore National Laboratory, CA (United States)

    1996-12-31T23:59:59.000Z

    On January 8, 1995, the Atomic Energy Organization of Iran (AEOI) signed a contract for $800 million with the Russian Federation Ministry for Atomic Energy (Minatom) to complete Bushehr nuclear power plant (BNPP) unit 1. The agreement called for a Russian VVER-1000/320 pressurized water reactor (PWR) to be successfully installed into the existing German-built BNPP facilities in 5 yr. System design differences, bomb damage, and environmental exposure are key issues with which Minatom must contend in order to fulfill the contract. The AEOI under the Shah of Iran envisioned Bushehr as the first of many nuclear power plants, with Iran achieving 24 GW(electric) by 1993 and 34 GW(electric) by 2000. Kraftwerk Union AG (KWU) began construction of the two-unit plant near the Persian Gulf town of Halileh in 1975. Unit 1 was {approx}80% complete and unit 2 was {approx}50% complete when construction was interrupted by the 1979 Iranian Islamic revolution. Despite repeated AEOI attempts to lure KWU and other companies back to Iran to complete the plant, Western concerns about nuclear proliferation in Iran and repeated bombings of the plant during the 1980-1988 Iran-Iraq war dissuaded Germany from resuming construction.

  1. Extending Sensor Calibration Intervals in Nuclear Power Plants

    SciTech Connect (OSTI)

    Coble, Jamie B.; Meyer, Ryan M.; Ramuhalli, Pradeep; Bond, Leonard J.; Shumaker, Brent; Hashemian, Hash

    2012-11-15T23:59:59.000Z

    Currently in the USA, sensor recalibration is required at every refueling outage, and it has emerged as a critical path item for shortening outage duration. International application of calibration monitoring, such as at the Sizewell B plant in UK, has shown that sensors may operate for eight years, or longer, within calibration tolerances. Online monitoring can be employed to identify those sensors which require calibration, allowing for calibration of only those sensors which need it. The US NRC accepted the general concept of online monitoring for sensor calibration monitoring in 2000, but no plants have been granted the necessary license amendment to apply it. This project addresses key issues in advanced recalibration methodologies and provides the science base to enable adoption of best practices for applying online monitoring, resulting in a public domain standardized methodology for sensor calibration interval extension. Research to develop this methodology will focus on three key areas: (1) quantification of uncertainty in modeling techniques used for calibration monitoring, with a particular focus on non-redundant sensor models; (2) accurate determination of acceptance criteria and quantification of the effect of acceptance criteria variability on system performance; and (3) the use of virtual sensor estimates to replace identified faulty sensors to extend operation to the next convenient maintenance opportunity.

  2. Guideline for the seismic technical evaluation of replacement items for nuclear power plants

    SciTech Connect (OSTI)

    Harris, S.P.; Cushing, R.W. (EQE International, San Francisco, CA (United States)); Johnson, H.W. (Programmatic Solutions, Smithtown, NY (United States)); Abeles, J.M. (System 1, Inc., Potomac, MD (United States))

    1993-02-01T23:59:59.000Z

    Seismic qualification for equipment originally installed in nuclear power plants was typically performed by the original equipment suppliers or manufactures (OES/OEM). Many of the OES/OEM no longer maintain quality assurance programs with adequate controls for supplying nuclear equipment. Utilities themselves must provide reasonable assurance in the continued seismic adequacy of such replacement items. This guideline provides practical, cost-effective techniques which can be used to provide reasonable assurance that replacement items will meet seismic performance requirements necessary to maintain the seismic design basis of commercial nuclear power plants. It also provides a method for determining when a seismic technical evaluation of replacement items (STERI) is required as part of the procurement process for spare and replacement items. Guidance on supplier program requirements necessary to maintain continued seismic adequacy and on documentation of maintaining required seismic adequacy is also included.

  3. State of the art review of radioactive waste volume reduction techniques for commercial nuclear power plants

    SciTech Connect (OSTI)

    Not Available

    1980-04-01T23:59:59.000Z

    A review is made of the state of the art of volume reduction techniques for low level liquid and solid radioactive wastes produced as a result of: (1) operation of commercial nuclear power plants, (2) storage of spent fuel in away-from-reactor facilities, and (3) decontamination/decommissioning of commercial nuclear power plants. The types of wastes and their chemical, physical, and radiological characteristics are identified. Methods used by industry for processing radioactive wastes are reviewed and compared to the new techniques for processing and reducing the volume of radioactive wastes. A detailed system description and report on operating experiences follow for each of the new volume reduction techniques. In addition, descriptions of volume reduction methods presently under development are provided. The Appendix records data collected during site surveys of vendor facilities and operating power plants. A Bibliography is provided for each of the various volume reduction techniques discussed in the report.

  4. The integrated workstation: A common, consistent link between nuclear plant personnel and plant information and computerized resources

    SciTech Connect (OSTI)

    Wood, R.T.; Knee, H.E.; Mullens, J.A.; Munro, J.K. Jr.; Swail, B.K.; Tapp, P.A.

    1993-01-01T23:59:59.000Z

    The increasing use of computer technology in the US nuclear power industry has greatly expanded the capability to obtain, analyze, and present data about the plant to station personnel. Data concerning a power plant's design, configuration, operational and maintenance histories, and current status, and the information that can be derived from them, provide the link between the plant and plant staff. It is through this information bridge that operations, maintenance and engineering personnel understand and manage plant performance. However, it is necessary to transform the vast quantity of data available from various computer systems and across communications networks into clear, concise, and coherent information. In addition, it is important to organize this information into a consolidated, structured form within an integrated environment so that various users throughout the plant have ready access at their local station to knowledge necessary for their tasks. Thus, integrated workstations are needed to provide the inquired information and proper software tools, in a manner that can be easily understood and used, to the proper users throughout the plant. An effort is underway at the Oak Ridge National Laboratory to address this need by developing Integrated Workstation functional requirements and implementing a limited-scale prototype demonstration. The integrated Workstation requirements will define a flexible, expandable computer environment that permits a tailored implementation of workstation capabilities and facilitates future upgrades to add enhanced applications. The functionality to be supported by the integrated workstation and inherent capabilities to be provided by the workstation environment win be described. In addition, general technology areas which are to be addressed in the Integrated Workstation functional requirements will be discussed.

  5. INTERNATIONAL JOURNAL OF HYDROGEN ENERGY Accepted June 2008 HYDROGEN STORAGE FOR MIXED WIND-NUCLEAR POWER PLANTS IN

    E-Print Network [OSTI]

    Cañizares, Claudio A.

    INTERNATIONAL JOURNAL OF HYDROGEN ENERGY Accepted June 2008 1 HYDROGEN STORAGE FOR MIXED WIND-NUCLEAR evaluation of hydrogen production and storage for a mixed wind-nuclear power plant considering some new of a combined nuclear-wind-hydrogen system is discussed first, where the selling and buying of electricity

  6. Decision to reorganise or reorganising decisions? A First-Hand Account of the Decommissioning of the Phnix Nuclear Power Plant

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    of the Decommissioning of the Phénix Nuclear Power Plant Melchior Pelleterat de Borde, MINES ParisTech, Christophe Martin prepared for decommissioning. This study, conducted between 2010 and 2012, is focused on the Phénix nuclear in the context of nuclear decommissioning. This article does not aim to present the results of the study, i

  7. Aging of steel containments and liners in nuclear power plants

    SciTech Connect (OSTI)

    Naus, D.J.; Oland, C.B. [Oak Ridge National Lab., TN (United States). Engineering Technology Div.; Ellingwood, B. [Johns Hopkins Univ., Baltimore, MD (United States). Dept. of Civil Engineering; Norris, W.E. [Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research

    1998-01-01T23:59:59.000Z

    Aging of the containment pressure boundary in light water reactor plants is being addressed to understand the significant factors relating occurrence of corrosion efficacy of inspection and structural capacity reduction of steel containments and liners of concrete containments. and to make recommendations on use of risk models in regulatory decisions. Current regulatory in-service inspection requirements are reviewed and a summary of containment related degradation experience is presented. Current and emerging nondestructive examination techniques and a degradation assessment methodology for characterizing and quantifying the amount of damage present are described. Quantitative tools for condition assessment of aging structures using time dependent structural reliability analysis methods are summarized. Such methods provide a framework for addressing the uncertainties attendant to aging in the decision process. Results of this research provide a means for establishing current and estimating future structural capacity margins of containments, and to address the significance of incidences of reported containment degradation.

  8. Diversity Strategies for Nuclear Power Plant Instrumentation and Control Systems

    SciTech Connect (OSTI)

    Wood, Richard Thomas [ORNL; Belles, Randy [ORNL; Cetiner, Mustafa Sacit [ORNL; Holcomb, David Eugene [ORNL; Korsah, Kofi [ORNL; Loebl, Andy [ORNL; Mays, Gary T [ORNL; Muhlheim, Michael David [ORNL; Mullens, James Allen [ORNL; Poore III, Willis P [ORNL; Qualls, A L [ORNL; Wilson, Thomas L [ORNL; Waterman, Michael E. [U.S. Nuclear Regulatory Commission

    2010-02-01T23:59:59.000Z

    This report presents the technical basis for establishing acceptable mitigating strategies that resolve diversity and defense-in-depth (D3) assessment findings and conform to U.S. Nuclear Regulatory Commission (NRC) requirements. The research approach employed to establish appropriate diversity strategies involves investigation of available documentation on D3 methods and experience from nuclear power and nonnuclear industries, capture of expert knowledge and lessons learned, determination of best practices, and assessment of the nature of common-cause failures (CCFs) and compensating diversity attributes. The research described in this report does not provide guidance on how to determine the need for diversity in a safety system to mitigate the consequences of potential CCFs. Rather, the scope of this report provides guidance to the staff and nuclear industry after a licensee or applicant has performed a D3 assessment per NUREG/CR-6303 and determined that diversity in a safety system is needed for mitigating the consequences of potential CCFs identified in the evaluation of the safety system design features. Succinctly, the purpose of the research described in this report was to answer the question, 'If diversity is required in a safety system to mitigate the consequences of potential CCFs, how much diversity is enough?' The principal results of this research effort have identified and developed diversity strategies, which consist of combinations of diversity attributes and their associated criteria. Technology, which corresponds to design diversity, is chosen as the principal system characteristic by which diversity criteria are grouped to form strategies. The rationale for this classification framework involves consideration of the profound impact that technology-focused design diversity provides. Consequently, the diversity usage classification scheme involves three families of strategies: (1) different technologies, (2) different approaches within the same technology, and (3) different architectures within the same technology. Using this convention, the first diversity usage family, designated Strategy A, is characterized by fundamentally diverse technologies. Strategy A at the system or platform level is illustrated by the example of analog and digital implementations. The second diversity usage family, designated Strategy B, is achieved through the use of distinctly different technologies. Strategy B can be described in terms of different digital technologies, such as the distinct approaches represented by general-purpose microprocessors and field-programmable gate arrays. The third diversity usage family, designated Strategy C, involves the use of variations within a technology. An example of Strategy C involves different digital architectures within the same technology, such as that provided by different microprocessors (e.g., Pentium and Power PC). The grouping of diversity criteria combinations according to Strategies A, B, and C establishes baseline diversity usage and facilitates a systematic organization of strategic approaches for coping with CCF vulnerabilities. Effectively, these baseline sets of diversity criteria constitute appropriate CCF mitigating strategies for digital safety systems. The strategies represent guidance on acceptable diversity usage and can be applied directly to ensure that CCF vulnerabilities identified through a D3 assessment have been adequately resolved. Additionally, a framework has been generated for capturing practices regarding diversity usage and a tool has been developed for the systematic assessment of the comparative effect of proposed diversity strategies (see Appendix A).

  9. Initial Scaling Studies and Conceptual Thermal Fluids Experiments for the Prismatic NGNP Point Design

    SciTech Connect (OSTI)

    D. M. McEligot; G. E. McCreery

    2004-09-01T23:59:59.000Z

    The objective of this report is to document the initial high temperature gas reactor scaling studies and conceptual experiment design for gas flow and heat transfer. The general approach of the project is to develop new benchmark experiments for assessment in parallel with CFD and coupled CFD/ATHENA/RELAP5-3D calculations for the same geometry. Two aspects of the complex flow in an NGNP are being addressed: (1) flow and thermal mixing in the lower plenum ("hot streaking" issue) and (2) turbulence and resulting temperature distributions in reactor cooling channels ("hot channel" issue). Current prismatic NGNP concepts are being examined to identify their proposed flow conditions and geometries over the range from normal operation to decay heat removal in a pressurized cooldown. Approximate analyses are being applied to determine key non-dimensional parameters and their magnitudes over this operating range. For normal operation, the flow in the coolant channels can be considered to be dominant forced convection with slight transverse property variation. The flow in the lower plenum can locally be considered to be a situation of multiple buoyant jets into a confined density-stratified crossflow -- with obstructions. Experiments are needed for the combined features of the lower plenum flows. Missing from the typical jet experiments are interactions with nearby circular posts and with vertical posts in the vicinity of vertical walls - with near stagnant surroundings at one extreme and significant crossflow at the other. Two heat transfer experiments are being considered. One addresses the "hot channel" problem, if necessary. The second experiment will treat heated jets entering a model plenum. Unheated MIR (Matched-Index-of-Refraction) experiments are first steps when the geometry is complicated. One does not want to use a computational technique which will not even handle constant properties properly. The MIR experiment will simulate flow features of the paths of jets as they mix in flowing through the array of posts in a lower plenum en route to the single exit duct. Initial conceptual designs for such experiments are described.

  10. Thermal Hydraulic Analyses for Coupling High Temperature Gas-Cooled Reactor to Hydrogen Plant

    SciTech Connect (OSTI)

    C.H. Oh; R. Barner; C. B. Davis; S. Sherman; P. Pickard

    2006-08-01T23:59:59.000Z

    The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood. A number of possible configurations for a system that transfers heat between the nuclear reactor and the hydrogen and/or electrical generation plants were identified. These configurations included both direct and indirect cycles for the production of electricity. Both helium and liquid salts were considered as the working fluid in the intermediate heat transport loop. Methods were developed to perform thermal-hydraulic and cycle-efficiency evaluations of the different configurations and coolants. The thermal-hydraulic evaluations estimated the sizes of various components in the intermediate heat transport loop for the different configurations. The relative sizes of components provide a relative indication of the capital cost associated with the various configurations. Estimates of the overall cycle efficiency of the various configurations were also determined. The evaluations determined which configurations and coolants are the most promising from thermalhydraulic and efficiency points of view.

  11. Auxiliary feedwater system risk-based inspection guide for the H. B. Robinson nuclear power plant

    SciTech Connect (OSTI)

    Moffitt, N.E.; Lloyd, R.C.; Gore, B.F.; Vo, T.V. [Pacific Northwest Lab., Richland, WA (United States); Garner, L.W. [Nuclear Regulatory Commission, Washington, DC (United States)

    1993-08-01T23:59:59.000Z

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. H. B. Robinson was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the H. B. Robinson plant.

  12. Auxiliary feedwater system risk-based inspection guide for the McGuire nuclear power plant

    SciTech Connect (OSTI)

    Bumgardner, J.D.; Lloyd, R.C.; Moffitt, N.E.; Gore, B.F.; Vo, T.V. [Pacific Northwest Lab., Richland, WA (United States)

    1994-05-01T23:59:59.000Z

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. McGuire was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the McGuire plant.

  13. Auxiliary feedwater system risk-based inspection guide for the J. M. Farley Nuclear Power Plant

    SciTech Connect (OSTI)

    Vo, T.V.; Pugh, R.; Gore, B.F.; Harrison, D.G. (Pacific Northwest Lab., Richland, WA (USA))

    1990-10-01T23:59:59.000Z

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment(PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. J. M. Farley was selected as the second plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important at the J. M. Farley plant. 23 refs., 1 fig., 1 tab.

  14. Auxiliary feedwater system risk-based inspection guide for the Ginna Nuclear Power Plant

    SciTech Connect (OSTI)

    Pugh, R.; Gore, B.F.; Vo, T.V.; Moffitt, N.E. (Pacific Northwest Lab., Richland, WA (United States))

    1991-09-01T23:59:59.000Z

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Ginna was selected as the eighth plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Ginna plant. 23 refs., 1 fig., 1 tab.

  15. Auxiliary feedwater system risk-based inspection guide for the South Texas Project nuclear power plant

    SciTech Connect (OSTI)

    Bumgardner, J.D.; Nickolaus, J.R.; Moffitt, N.E.; Gore, B.F.; Vo, T.V. [Pacific Northwest Lab., Richland, WA (United States)

    1993-12-01T23:59:59.000Z

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. South Texas Project was selected as a plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by the NRC inspectors in preparation of inspection plans addressing AFW risk important components at the South Texas Project plant.

  16. Integrated head package cable carrier for a nuclear power plant

    DOE Patents [OSTI]

    Meuschke, Robert E. (Monroeville, PA); Trombola, Daniel M. (Murrysville, PA)

    1995-01-01T23:59:59.000Z

    A cabling arrangement is provided for a nuclear reactor located within a containment. Structure inside the containment is characterized by a wall having a near side surrounding the reactor vessel defining a cavity, an operating deck outside the cavity, a sub-space below the deck and on a far side of the wall spaced from the near side, and an operating area above the deck. The arrangement includes a movable frame supporting a plurality of cables extending through the frame, each connectable at a first end to a head package on the reactor vessel and each having a second end located in the sub-space. The frame is movable, with the cables, between a first position during normal operation of the reactor when the cables are connected to the head package, located outside the sub-space proximate the head package, and a second position during refueling when the cables are disconnected from the head package, located in the sub-space. In a preferred embodiment, the frame straddles the top of the wall in a substantially horizontal orientation in the first position, pivots about an end distal from the head package to a substantially vertically oriented intermediate position, and is guided, while remaining about vertically oriented, along a track in the sub-space to the second position.

  17. Development of a checklist for evaluating emergency procedures used in nuclear power plants

    SciTech Connect (OSTI)

    Brune, R.L.; Weinstein, M.

    1981-05-01T23:59:59.000Z

    This report describes the process for developing a checklist to be used by US Nuclear Regulatory Commission Office of Inspection and Enforcement (I and E) inspectors during their evaluation of emergency procedures used in nuclear power plants. The objective of the checklist is to aid inspectors in identifying procedural characteristics that can lead to reactor operator performance deviations. Four nuclear power plants were surveyed to obtain a sample of procedures and related information for human factors evaluation. In addition, a human factors analysis of 890 LERs submitted during the period 1975 through 1978 was performed to identify the major categories of performance deviations associated with reactor operator activities. Checklist items aimed at preventing these performance deviations or facilitating their early detection were developed. The study findings supporting the procedures evaluation criteria comprising the checklist items are described in this report. A companion document, Checklist for Evaluating Emergency Procedures Used in Nuclear Power Plants, NUREG/CR-2005, SAND81-7074, has been prepared as a handbook for inspectors. It describes the checklist and provides instructions for its use. 24 figs.

  18. Method of installing a control room console in a nuclear power plant

    DOE Patents [OSTI]

    Scarola, Kenneth (Windsor, CT); Jamison, David S. (Windsor, CT); Manazir, Richard M. (North Canton, CT); Rescorl, Robert L. (Vernon, CT); Harmon, Daryl L. (Enfield, CT)

    1994-01-01T23:59:59.000Z

    An advanced control room complex for a nuclear power plant, including a discrete indicator and alarm system (72) which is nuclear qualified for rapid response to changes in plant parameters and a component control system (64) which together provide a discrete monitoring and control capability at a panel (14-22, 26, 28) in the control room (10). A separate data processing system (70), which need not be nuclear qualified, provides integrated and overview information to the control room and to each panel, through CRTs (84) and a large, overhead integrated process status overview board (24). The discrete indicator and alarm system (72) and the data processing system (70) receive inputs from common plant sensors and validate the sensor outputs to arrive at a representative value of the parameter for use by the operator during both normal and accident conditions, thereby avoiding the need for him to assimilate data from each sensor individually. The integrated process status board (24) is at the apex of an information hierarchy that extends through four levels and provides access at each panel to the full display hierarchy. The control room panels are preferably of a modular construction, permitting the definition of inputs and outputs, the man machine interface, and the plant specific algorithms, to proceed in parallel with the fabrication of the panels, the installation of the equipment and the generic testing thereof.

  19. Development, Application, and Implementation of RAMCAP to Characterize Nuclear Power Plant Risk From Terrorism

    SciTech Connect (OSTI)

    Gaertner, John P. [Electric Power Research Institute, 1300 Harris Boulevard, Charlotte, NC 28262 (United States); Teagarden, Grant A. [ERIN Engineering and Research (United States)

    2006-07-01T23:59:59.000Z

    In response to increased interest in risk-informed decision making regarding terrorism, EPRI and ERIN Engineering were selected by U.S. DHS and ASME to develop and demonstrate the RAMCAP method for nuclear power plant (NPP) risk assessment. The objective is to characterize plant-specific NPP risk for risk management opportunities and to provide consistent information for DHS decision making. This paper is an update of this project presented at the American Nuclear Society (ANS) International Topical Meeting on Probabilistic Safety Analysis (PSA05) in September, 2005. The method uses a characterization of risk as a function of Consequence, Vulnerability, and Threat. For each site, worst case scenarios are developed for each of sixteen benchmark threats. Nuclear RAMCAP hypothesizes that the intent of the perpetrator is to cause offsite radiological consequences. Specific targets are the reactor core, the spent fuel pool, and nuclear spent fuel in a dry storage facility (ISFSI). Results for each scenario are presented as conditional risk for financial loss, early fatalities and early injuries. Expected consequences for each scenario are quantified, while vulnerability is estimated on a relative likelihood scale. Insights for other societal risks are provided. Although threat frequencies are not provided, target attractiveness and threat deterrence are estimated. To assure efficiency, completeness, and consistency; results are documented using standard RAMCAP Evaluator software. Trial applications were successfully performed at four plant sites. Implementation at all other U.S. commercial sites is underway, supported by the Nuclear Sector Coordinating Council (NSCC). Insights from RAMCAP results at 23 U.S. plants completed to date have been compiled and presented to the NSCC. Results are site-specific. Physical security barriers, an armed security force, preparedness for design-basis threats, rugged design against natural hazards, multiple barriers between fuel and environment, accident mitigation capability, severe accident management procedures, and offsite emergency plans are risk-beneficial against all threat types. (authors)

  20. U.S. Seafood Safe and Unaffected by Radiation Contamination from Japanese Nuclear Power Plant Incident; U.S. Monitoring Control Strategy Explained

    E-Print Network [OSTI]

    U.S. Seafood Safe and Unaffected by Radiation Contamination from Japanese Nuclear Power Plant about radiation contamination from the Japanese nuclear power plant incident and on the control potential routes by which seafood contaminated with radionuclides from the Japanese nuclear power plant

  1. Redundant Sensor Calibration and Estimation for Monitoring and Control of Nuclear Power Plants Xin Jin, Asok Ray and Robert M. Edwards

    E-Print Network [OSTI]

    Ray, Asok

    Redundant Sensor Calibration and Estimation for Monitoring and Control of Nuclear Power Plants Xin@engr.psu.edu INTRODUCTION Performance, reliability and safety of nuclear power plants depend upon validity and accuracy are installed with redundancy in nuclear power plants. Redundancy can be classified into two groups: direct

  2. Power to the People or Regulatory Ratcheting? Explaining the Success (or Failure) of Attempts to Site Commercial U.S. Nuclear Power Plants: 1954 -19961

    E-Print Network [OSTI]

    to Site Commercial U.S. Nuclear Power Plants: 1954 - 19961 7 April 2014 Eric Berndt2 and Daniel P. Aldrich to attempt siting nuclear power plant facilities in large numbers in the 1960s. By the late 1990s, more than 1984). In the case of the Shoreham Nuclear Generating Station in Long Island, the plant was completed

  3. The Handbook of Applied Bayesian Analysis, Eds: Tony O'Hagan & Mike West, Oxford University Bayesian analysis and decisions in nuclear power plant

    E-Print Network [OSTI]

    Morton, David

    Bayesian analysis and decisions in nuclear power plant maintenance Elmira Popova, David Morton, Paul Damien are then applied to solving an important problem in a nuclear power plant system at the South Texas Project (STP) Electric Generation Station. STP is one of the newest and largest nuclear power plants in the US

  4. Applicability of Operational Research Techniques in CANDU Nuclear Plant Maintenance

    SciTech Connect (OSTI)

    Doyle, E. Kevin [Bruce Power LP, Box 4000B12, Tiverton, Ont., N0G2T0 (Canada)

    2002-07-01T23:59:59.000Z

    As previously reported at ICONE 6 in New Orleans, 1996, and ICONE 9 in Niece, 2001, the use of various maintenance optimization techniques at Bruce has lead to cost effective preventive maintenance applications for complex systems. Innovative practices included greatly reducing Reliability Centered Maintenance (RCM) costs while maintaining the accuracy of the analysis. The optimization strategy has undergone further evolution and at the present an Integrated Maintenance Program (IMP) is being put in place. Further cost refinement of the station preventive maintenance strategy whereby decisions are based on statistical analysis of historical failure data is being evaluated. A wide range of Operational Research (OR) literature was reviewed for implementation issues and several encouraging areas were found that will assist in the current effort of evaluating maintenance optimization techniques for nuclear power production. The road ahead is expected to consist first of resolving 25 years of data issues and preserving the data via appropriate knowledge system techniques while post war demographics permit experts to input into the system. Subsequent analytical techniques will emphasize total simplicity to obtain the requisite buy in from Corporate Executives who possibly are not trained in Operational Research. Case studies of containment airlock seal failures are used to illustrate the direct applicability of stochastic processes. Airlocks and transfer chambers were chosen as they have long been known as high maintenance items. Also, the very significant financial consequences of this type of failure will help to focus the attention of Senior Management on the effort. Despite substantial investment in research, improvement in the design of the seal material or configuration has not been achieved beyond the designs completed in the 1980's. Overall, the study showed excellent agreement of the relatively quick stochastic methods with the maintenance programs produced at great cost over years of trial and error. The pivotal role of expert opinion via experienced users/problem owners/maintenance engineers in all phases of the method and its application was noted and will be explored in subsequent efforts. The results are displayed via economic alternatives to more easily attract the attention of Maintenance Managers. Graphical overviews of the data demonstrated that substantial insight can be gained by simply organizing the data into statistically meaningful arrays such as histograms. The conclusions highlight several very positive avenues to evaluate at this particular juncture in time. (author)

  5. Space Nuclear Power Plant Pre-Conceptual Design Report, For Information

    SciTech Connect (OSTI)

    B. Levine

    2006-01-27T23:59:59.000Z

    This letter transmits, for information, the Project Prometheus Space Nuclear Power Plant (SNPP) Pre-Conceptual Design Report completed by the Naval Reactors Prime Contractor Team (NRPCT). This report documents the work pertaining to the Reactor Module, which includes integration of the space nuclear reactor with the reactor radiation shield, energy conversion, and instrumentation and control segments. This document also describes integration of the Reactor Module with the Heat Rejection segment, the Power Conditioning and Distribution subsystem (which comprise the SNPP), and the remainder of the Prometheus spaceship.

  6. Considerations in the evaluation of concrete structures for continued service in aged Nuclear Power Plants (NPPs)

    SciTech Connect (OSTI)

    Naus, D.; Marchbanks, M.; Oland, B.; Arndt, G.; Brown, T.

    1989-01-01T23:59:59.000Z

    Currently, there are /approximately/119 commercial nuclear power plants (NPPs) in the US either under construction, operating at low-to-full power, or awaiting an operating license. Together, these units have a net generating capacity of /approximately/110 GW(e). Assuming no life extension of present facilities, the operating licenses for these plants will start to expire in the middle of the next decade with Yankee Rowe being the first plant to attain this status. Where it is noted that with no life extension of facilities, a potential loss of electrical generating capacity in excess of 75 GW(e) could occur during the time period 2006 to 2020 when the operating licenses of 80 to 90 NPPs are scheduled to expire. A potential timely and cost-effective solution to meeting future electricity demand, which has worked well for non-nuclear generating plants, is to extend the service life (operating licenses) of existing NPPs. Since the concrete components in these plants provide a vital safety function, any continued service considerations must include an in-depth assessment of the safety-related concrete structures. 7 refs.

  7. Impact of structural aging on seismic risk assessment of reinforced concrete structures in nuclear power plants

    SciTech Connect (OSTI)

    Ellingwood, B.; Song, J. [Johns Hopkins Univ., Baltimore, MD (United States). Dept. of Civil Engineering

    1996-03-01T23:59:59.000Z

    The Structural Aging Program is addressing the potential for degradation of concrete structural components and systems in nuclear power plants over time due to aging and aggressive environmental stressors. Structures are passive under normal operating conditions but play a key role in mitigating design-basis events, particularly those arising from external challenges such as earthquakes, extreme winds, fires and floods. Structures are plant-specific and unique, often are difficult to inspect, and are virtually impossible to replace. The importance of structural failures in accident mitigation is amplified because such failures may lead to common-cause failures of other components. Structural condition assessment and service life prediction must focus on a few critical components and systems within the plant. Components and systems that are dominant contributors to risk and that require particular attention can be identified through the mathematical formalism of a probabilistic risk assessment, or PRA. To illustrate, the role of structural degradation due to aging on plant risk is examined through the framework of a Level 1 seismic PRA of a nuclear power plant. Plausible mechanisms of structural degradation are found to increase the core damage probability by approximately a factor of two.

  8. France gets nuclear fusion plant France will get to host the project to build a 10bn-euro (6.6bn) nuclear fusion reactor, in

    E-Print Network [OSTI]

    ) nuclear fusion reactor, in the face of strong competition from Japan. The International Thermonuclear division, which is responsible for the UK's thermonuclear fusion programme, said the decisionFrance gets nuclear fusion plant France will get to host the project to build a 10bn-euro (£6.6bn

  9. Potential safety-related incidents with possible applicability to a nuclear fuel reprocessing plant

    SciTech Connect (OSTI)

    Durant, W.S.; Perkins, W.C.; Lee, R.; Stoddard, D.H.

    1982-05-20T23:59:59.000Z

    The Safety Technology Group is developing methodology that can be used to assess the risk of operating a plant to reprocess spent nuclear fuel. As an early step in the methodology, a preliminary hazards analysis identifies safety-related incidents. In the absence of appropriate safety features, these incidents could lead to significant consequences and risk to onsite personnel or to the public. This report is a compilation of potential safety-related incidents that have been identified in studies at SRL and in safety analyses of various commercially designed reprocessing plants. It is an expanded revision of the version originally published as DP-1558, Published December 1980.

  10. Auxiliary feedwater system risk-based inspection guide for the Byron and Braidwood nuclear power plants

    SciTech Connect (OSTI)

    Moffitt, N.E.; Gore, B.F.: Vo, T.V. (Pacific Northwest Lab., Richland, WA (USA))

    1991-07-01T23:59:59.000Z

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Byron and Braidwood were selected for the fourth study in this program. The produce of this effort is a prioritized listing of AFW failures which have occurred at the plants and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Byron/Braidwood plants. 23 refs., 1 fig., 1 tab.

  11. Uncertainty Quantification Techniques for Sensor Calibration Monitoring in Nuclear Power Plants

    SciTech Connect (OSTI)

    Ramuhalli, Pradeep; Lin, Guang; Crawford, Susan L.; Konomi, Bledar A.; Coble, Jamie B.; Shumaker, Brent; Hashemian, Hash

    2014-04-30T23:59:59.000Z

    This report describes research towards the development of advanced algorithms for online calibration monitoring. The objective of this research is to develop the next generation of online monitoring technologies for sensor calibration interval extension and signal validation in operating and new reactors. These advances are expected to improve the safety and reliability of current and planned nuclear power systems as a result of higher accuracies and increased reliability of sensors used to monitor key parameters. The focus of this report is on documenting the outcomes of the first phase of R&D under this project, which addressed approaches to uncertainty quantification (UQ) in online monitoring that are data-driven, and can therefore adjust estimates of uncertainty as measurement conditions change. Such data-driven approaches to UQ are necessary to address changing plant conditions, for example, as nuclear power plants experience transients, or as next-generation small modular reactors (SMR) operate in load-following conditions.

  12. Uncertainty Quantification Techniques for Sensor Calibration Monitoring in Nuclear Power Plants

    SciTech Connect (OSTI)

    Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Lin, Guang [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Crawford, Susan L. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Konomi, Bledar A. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Braatz, Brett G. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Coble, Jamie B. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Shumaker, Brent [Analysis and Measurement Services Corp., Knoxville, TN (United States); Hashemian, Hash [Analysis and Measurement Services Corp., Knoxville, TN (United States)

    2013-09-01T23:59:59.000Z

    This report describes the status of ongoing research towards the development of advanced algorithms for online calibration monitoring. The objective of this research is to develop the next generation of online monitoring technologies for sensor calibration interval extension and signal validation in operating and new reactors. These advances are expected to improve the safety and reliability of current and planned nuclear power systems as a result of higher accuracies and increased reliability of sensors used to monitor key parameters. The focus of this report is on documenting the outcomes of the first phase of R&D under this project, which addressed approaches to uncertainty quantification (UQ) in online monitoring that are data-driven, and can therefore adjust estimates of uncertainty as measurement conditions change. Such data-driven approaches to UQ are necessary to address changing plant conditions, for example, as nuclear power plants experience transients, or as next-generation small modular reactors (SMR) operate in load-following conditions.

  13. Devices and methods for managing noncombustible gasses in nuclear power plants

    DOE Patents [OSTI]

    Marquino, Wayne; Moen, Stephan C; Wachowiak, Richard M; Gels, John L; Diaz-Quiroz, Jesus; Burns, Jr., John C

    2014-12-23T23:59:59.000Z

    Systems passively eliminate noncondensable gasses from facilities susceptible to damage from combustion of built-up noncondensable gasses, such as H2 and O2 in nuclear power plants, without the need for external power and/or moving parts. Systems include catalyst plates installed in a lower header of the Passive Containment Cooling System (PCCS) condenser, a catalyst packing member, and/or a catalyst coating on an interior surface of a condensation tube of the PCCS condenser or an annular outlet of the PCCS condenser. Structures may have surfaces or hydrophobic elements that inhibit water formation and promote contact with the noncondensable gas. Noncondensable gasses in a nuclear power plant are eliminated by installing and using the systems individually or in combination. An operating pressure of the PCCS condenser may be increased to facilitate recombination of noncondensable gasses therein.

  14. A survey of repair practices for nuclear power plant containment metallic pressure boundaries

    SciTech Connect (OSTI)

    Oland, C.B.; Naus, D.J. [Oak Ridge National Lab., TN (United States)

    1998-05-01T23:59:59.000Z

    The Nuclear Regulatory Commission has initiated a program at the Oak Ridge National Laboratory to provide assistance in their assessment of the effects of potential degradation on the structural integrity and leaktightness of metal containment vessels and steel liners of concrete containments in nuclear power plants. One of the program objectives is to identify repair practices for restoring metallic containment pressure boundary components that have been damaged or degraded in service. This report presents issues associated with inservice condition assessments and continued service evaluations and identifies the rules and requirements for the repair and replacement of nonconforming containment pressure boundary components by welding or metal removal. Discussion topics include base and welding materials, welding procedure and performance qualifications, inspection techniques, testing methods, acceptance criteria, and documentation requirements necessary for making acceptable repairs and replacements so that the plant can be returned to a safe operating condition.

  15. Handbook of nuclear power plant seismic fragilities, Seismic Safety Margins Research Program

    SciTech Connect (OSTI)

    Cover, L.E.; Bohn, M.P.; Campbell, R.D.; Wesley, D.A.

    1983-12-01T23:59:59.000Z

    The Seismic Safety Margins Research Program (SSMRP) has a gola to develop a complete fully coupled analysis procedure (including methods and computer codes) for estimating the risk of an earthquake-induced radioactive release from a commercial nuclear power plant. As part of this program, calculations of the seismic risk from a typical commercial nuclear reactor were made. These calculations required a knowledge of the probability of failure (fragility) of safety-related components in the reactor system which actively participate in the hypothesized accident scenarios. This report describes the development of the required fragility relations and the data sources and data reduction techniques upon which they are based. Both building and component fragilities are covered. The building fragilities are for the Zion Unit 1 reactor which was the specific plant used for development of methodology in the program. Some of the component fragilities are site-specific also, but most would be usable for other sites as well.

  16. Summary and analysis of public comments on NUREG-1317: Regulatory options for nuclear plant license renewal: Final report

    SciTech Connect (OSTI)

    Ligon, D.M.; Seth, S.S.

    1989-03-01T23:59:59.000Z

    On August 29, 1988, the US Nuclear Regulatory Commission (NRC) issued an Advance Notice of Proposed Rulemaking on nuclear plant license renewal and solicited public comments on NUREG-1317, ''Regulatory Options for Nuclear Plant License Renewal.'' NUREG-1317 presents a discussion of fifteen topics involving technical, environmental, and procedural issues and poses a set of related questions. As part of its ongoing task for the NRC, The MITRE Corporation has summarized and analyzed the public comments received. Fifty-three written comments were received. Of these, 83 percent were from nuclear industry representatives; the remaining comments represented federal and state agencies, public interest groups, and a private citizen.

  17. The History and Future of NDE in the Management of Nuclear Power Plant Materials Degradation

    SciTech Connect (OSTI)

    Doctor, Steven R.

    2009-04-01T23:59:59.000Z

    The author has spent more than 25 years conducting engineering and research studies to quantify the performance of nondestructive evaluation (NDE) in nuclear power plant (NPP) applications and identifying improvements to codes and standards for NDE to manage materials degradation. This paper will review this fundamental NDE engineering/research work and then look to the future on how NDE can be optimized for proactively managing materials degradation in NPP components.

  18. Dose-projection considerations for emergency conditions at nuclear power plants

    SciTech Connect (OSTI)

    Stoetzel, G.A.; Ramsdell, J.V.; Poeton, R.W.; Powell, D.C.; Desrosiers, A.E.

    1983-05-01T23:59:59.000Z

    The purpose of this report is to review the problems and issues associated with making environmental radiation-dose projections during emergencies at nuclear power plants. The review is divided into three areas: source-term development, characterization of atmospheric dispersion and selection of appropriate dispersion models, and development of dosimetry calculations for determining thyroid dose and whole-body dose for ground-level and elevated releases. A discussion of uncertainties associated with these areas is also provided.

  19. Separation Requirements for a Hydrogen Production Plant and High-Temperature Nuclear Reactor

    SciTech Connect (OSTI)

    Curtis Smith; Scott Beck; Bill Galyean

    2005-09-01T23:59:59.000Z

    This report provides the methods, models, and results of an evaluation for locating a hydrogen production facility near a nuclear power plant. In order to answer the risk-related questions for this combined nuclear and chemical facility, we utilized standard probabilistic safety assessment methodologies to answer three questions: what can happen, how likely is it, and what are the consequences? As part of answering these questions, we developed a model suitable to determine separation distances for hydrogen process structures and the nuclear plant structures. Our objective of the model-development and analysis is to answer key safety questions related to the placement of one or more hydrogen production plants in the vicinity of a high-temperature nuclear reactor. From a thermal-hydraulic standpoint we would like the two facilities to be quite close. However, safety and regulatory implications force the separation distance to be increased, perhaps substantially. Without answering these safety questions, the likelihood for obtaining a permit to construct and build such as facility in the U.S. would be questionable. The quantitative analysis performed for this report provides us with a scoping mechanism to determine key parameters related to the development of a nuclear-based hydrogen production facility. From our calculations, we estimate that when the separation distance is less than 100m, the core damage frequency is large enough (greater than 1E-6/yr) to become problematic in a risk-informed environment. However, a variety of design modifications, for example blast-deflection barriers, were explored to determine the impact of potential mitigating strategies. We found that these mitigating cases may significantly reduce risk and should be explored as the design for the hydrogen production facility evolves.

  20. Reviewing PSA-based analyses to modify technical specifications at nuclear power plants

    SciTech Connect (OSTI)

    Samanta, P.K.; Martinez-Guridi, G. [Brookhaven National Lab., Upton, NY (United States); Vesely, W.E. [Science Applications International Corporation, Dublin, OH (United States)

    1995-12-01T23:59:59.000Z

    Changes to Technical Specifications (TSs) at nuclear power plants (NPPs) require review and approval by the United States Nuclear Regulatory Commission (USNRC). Currently, many requests for changes to TSs use analyses that are based on a plant`s probabilistic safety assessment (PSA). This report presents an approach to reviewing such PSA-based submittals for changes to TSs. We discuss the basic objectives of reviewing a PSA-based submittal to modify NPP TSs; the methodology of reviewing a TS submittal, and the differing roles of a PSA review, a PSA Computer Code review, and a review of a TS submittal. To illustrate this approach, we discuss our review of changes to allowed outage time (AOT) and surveillance test interval (STI) in the TS for the South Texas Project Nuclear Generating Station. Based on this experience gained, a check-list of items is given for future reviewers; it can be used to verify that the submittal contains sufficient information, and also that the review has addressed the relevant issues. Finally, recommended steps in the review process and the expected findings of each step are discussed.

  1. Data base on dose reduction research projects for nuclear power plants. Volume 5

    SciTech Connect (OSTI)

    Khan, T.A.; Yu, C.K.; Roecklein, A.K. [Brookhaven National Lab., Upton, NY (United States)] [Brookhaven National Lab., Upton, NY (United States)

    1994-05-01T23:59:59.000Z

    This is the fifth volume in a series of reports that provide information on dose reduction research and health physics technology or nuclear power plants. The information is taken from two of several databases maintained by Brookhaven National Laboratory`s ALARA Center for the Nuclear Regulatory Commission. The research section of the report covers dose reduction projects that are in the experimental or developmental phase. It includes topics such as steam generator degradation, decontamination, robotics, improvements in reactor materials, and inspection techniques. The section on health physics technology discusses dose reduction efforts that are in place or in the process of being implemented at nuclear power plants. A total of 105 new or updated projects are described. All project abstracts from this report are available to nuclear industry professionals with access to a fax machine through the ACEFAX system or a computer with a modem and the proper communications software through the ACE system. Detailed descriptions of how to access all the databases electronically are in the appendices of the report.

  2. Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

    SciTech Connect (OSTI)

    NONE

    1995-08-01T23:59:59.000Z

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized into six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.

  3. AGE-RELATED DEGRADATION OF NUCLEAR POWER PLANT STRUCTURES AND COMPONENTS.

    SciTech Connect (OSTI)

    BRAVERMAN,J.

    1999-03-29T23:59:59.000Z

    This paper summarizes and highlights the results of the initial phase of a research project on the assessment of aged and degraded structures and components important to the safe operation of nuclear power plants (NPPs). A review of age-related degradation of structures and passive components at NPPs was performed. Instances of age-related degradation have been collected and reviewed. Data were collected from plant generated documents such as Licensing Event Reports, NRC generic communications, NUREGs and industry reports. Applicable cases of degradation occurrences were reviewed and then entered into a computerized database. The results obtained from the review of degradation occurrences are summarized and discussed. Various trending analyses were performed to identify which structures and components are most affected, whether degradation occurrences are worsening, and what are the most common aging mechanisms. The paper also discusses potential aging issues and degradation-susceptible structures and passive components which would have the greatest impact on plant risk.

  4. Aging management guideline for commercial nuclear power plants-stationary batteries. Final report

    SciTech Connect (OSTI)

    Berg, R.; Shao, J.; Krencicki, G.; Giachetti, R. [Multiple Dynamics Corp., Southfield, MI (United States)

    1994-03-01T23:59:59.000Z

    The Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant stationary batteries important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.

  5. Age-Related Degradation of Nuclear Power Plant Structures and Components

    SciTech Connect (OSTI)

    Braverman, J.; Chang, T.-Y.; Chokshi, N.; Hofmayer, C.; Morante, R.; Shteyngart, S.

    1999-03-29T23:59:59.000Z

    This paper summarizes and highlights the results of the initial phase of a research project on the assessment of aged and degraded structures and components important to the safe operation of nuclear power plants (NPPs). A review of age-related degradation of structures and passive components at NPPs was performed. Instances of age-related degradation have been collected and reviewed. Data were collected from plant generated documents such as Licensing Event Reports, NRC generic communications, NUREGs and industry reports. Applicable cases of degradation occurrences were reviewed and then entered into a computerized database. The results obtained from the review of degradation occurrences are summarized and discussed. Various trending analyses were performed to identify which structures and components are most affected, whether degradation occurrences are worsening, and what was the most common aging mechanisms. The paper also discusses potential aging issues and degradation-susceptible structures and passive components which would have the greatest impact on plant risk.

  6. Aging Management Guideline for commercial nuclear power plants: Electrical switchgear. Final report

    SciTech Connect (OSTI)

    Toman, G.; Gazdzinski, R.; Schuler, K. [Ogden Environmental and Energy Services Co., Inc., Blue Bell, PA (United States)

    1993-07-01T23:59:59.000Z

    This Aging Management Guideline (AMG) provides recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant electrical switchgear important to license renewal. The latent of this AMG to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner which allows personnel responsible for performance analysis and maintenance, to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.

  7. Assessment of the radiological impact of a decommissioning nuclear power plant in Italy

    E-Print Network [OSTI]

    A. Petraglia; C. Sabbarese; M. De Cesare; N. De Cesare; F. Quinto; F. Terrasi; A. D'Onofrio; P. Steier; L. K. Fifield; A. M. Esposito

    2012-07-17T23:59:59.000Z

    The assessment of the radiological impact of a decommissioning Nuclear Power Plant is presented here through the results of an environmental monitoring survey carried out in the area surrounding the Garigliano Power Plant. The levels of radioactivity in soil, water, air and other environmental matrices are shown, in which {\\alpha}, {\\beta} and {\\gamma} activity and {\\gamma} equivalent dose rate are measured. Radioactivity levels of the samples from the Garigliano area are analyzed and then compared to those from a control zone situated more than 100 km away. Moreover, a comparison is made with a previous survey held in 2001. The analyses and comparisons show no significant alteration in the radiological characteristics of the area surroundings the plant, with an overall radioactivity depending mainly from the global fallout and natural sources.

  8. Assessment of the radiological impact of a decommissioning nuclear power plant in Italy

    E-Print Network [OSTI]

    Petraglia, A; De Cesare, M; De Cesare, N; Quinto, F; Terrasi, F; D'Onofrio, A; Steier, P; Fifield, L K; Esposito, A M; 10.1051/radiopro/2012010

    2012-01-01T23:59:59.000Z

    The assessment of the radiological impact of a decommissioning Nuclear Power Plant is presented here through the results of an environmental monitoring survey carried out in the area surrounding the Garigliano Power Plant. The levels of radioactivity in soil, water, air and other environmental matrices are shown, in which {\\alpha}, {\\beta} and {\\gamma} activity and {\\gamma} equivalent dose rate are measured. Radioactivity levels of the samples from the Garigliano area are analyzed and then compared to those from a control zone situated more than 100 km away. Moreover, a comparison is made with a previous survey held in 2001. The analyses and comparisons show no significant alteration in the radiological characteristics of the area surroundings the plant, with an overall radioactivity depending mainly from the global fallout and natural sources.

  9. Numerical simulation of the thermal conditions in a sea bay water area used for water supply to nuclear power plants

    SciTech Connect (OSTI)

    Sokolov, A. S. [JSC 'B. E. Vedeneev All-Russia Research Institute of Hydraulic Engineering (VNIIG)' (Russian Federation)] [JSC 'B. E. Vedeneev All-Russia Research Institute of Hydraulic Engineering (VNIIG)' (Russian Federation)

    2013-07-15T23:59:59.000Z

    Consideration is given to the numerical simulation of the thermal conditions in sea water areas used for both water supply to and dissipation of low-grade heat from a nuclear power plant on the shore of a sea bay.

  10. Risk-informed public safety policy for seismic events in the vicinity of a nuclear power plant

    E-Print Network [OSTI]

    Afolayan Jejeloye, Olubukola

    2002-01-01T23:59:59.000Z

    Nuclear Power Plants (NPPs) are potentially vulnerable to accidents, which can either be internally or externally initiated. External events include natural events like tornadoes, hurricanes, and earthquakes. The purpose ...

  11. Comparative analysis of United States and French nuclear power plant siting and construction regulatory policies and their economic consequences

    E-Print Network [OSTI]

    Golay, Michael Warren.

    1977-01-01T23:59:59.000Z

    Despite the substantial commitments of time and money which are devoted to the nuclear power plant siting process, the effectiveness of the system in providing a balanced evaluation of the technical, environmental and ...

  12. The potential role of new technology for enhanced safety and performance of nuclear power plants through improved service maintenance

    E-Print Network [OSTI]

    Achorn, Ted Glen

    1991-01-01T23:59:59.000Z

    Refinements in the safety and performance of nuclear power plants must be made to maintain public confidence and ensure competitiveness with other power sources. The aircraft industry, US Navy, and other programs have ...

  13. Development of a hybrid intelligent system for on-line real-time monitoring of nuclear power plant operations

    E-Print Network [OSTI]

    Yildiz, Bilge, 1976-

    2003-01-01T23:59:59.000Z

    A nuclear power plant (NPP) has an intricate operational domain involving systems, structures and components (SSCs) that vary in scale and complexity. Many of the large scale SSCs contribute to the lost availability in the ...

  14. Economic Analysis of a Nuclear Reactor Powered High-Temperature Electrolysis Hydrogen Production Plant

    SciTech Connect (OSTI)

    E. A. Harvego; M. G. McKellar; M. S. Sohal; J. E. O'Brien; J. S. Herring

    2008-08-01T23:59:59.000Z

    A reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production was developed to provide a basis for comparing the HTE concept with other hydrogen production concepts. The reference plant design is driven by a high-temperature helium-cooled nuclear reactor coupled to a direct Brayton power cycle. The reference design reactor power is 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 540°C and 900°C, respectively. The electrolysis unit used to produce hydrogen includes 4,009,177 cells with a per-cell active area of 225 cm2. The optimized design for the reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes an air-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The inlet air for the air-sweep system is compressed to the system operating pressure of 5.0 MPa in a four-stage compressor with intercooling. The alternating-current, AC, to direct-current, DC, conversion efficiency is 96%. The overall system thermal-to-hydrogen production efficiency (based on the lower heating value of the produced hydrogen) is 47.12% at a hydrogen production rate of 2.356 kg/s. An economic analysis of this plant was performed using the standardized H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program, and using realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.23/kg of hydrogen was calculated assuming an internal rate of return of 10%.

  15. Next Generation Nuclear Plant Project Technology Development Roadmaps: The Technical Path Forward for 750–800°C Reactor Outlet Temperature

    SciTech Connect (OSTI)

    John Collins

    2009-08-01T23:59:59.000Z

    This document presents the NGNP Critical PASSCs and defines their technical maturation path through Technology Development Roadmaps (TDRMs) and their associated Technology Readiness Levels (TRLs). As the critical PASSCs advance through increasing levels of technical maturity, project risk is reduced and the likelihood of within-budget and on-schedule completion is enhanced. The current supplier-generated TRLs and TDRMs for a 750–800°C reactor outlet temperature (ROT) specific to each supplier are collected in Appendix A.

  16. Evaluation criteria and procedure for nuclear power plant temporary loads/temporary conditions

    SciTech Connect (OSTI)

    Tang, H.T. [Electric Power Research Inst., Palo Alto, CA (United States); Minichiello, J.C. [Commonwealth Edison Co., Downers Grove, IL (United States); Olson, D.E. [Sargent and Lundy, Chicago, IL (United States)

    1996-12-01T23:59:59.000Z

    Operating nuclear power plants frequently encounter temporary loads/temporary conditions in plant normal operation and maintenance (O and M). The most obvious examples are installation of temporary shielding and scaffolding, or removal of certain supports, to facilitate plant refueling and maintenance outage activities. Short-term operability calls such as those due to snubber failures or unanticipated transients also create temporary loads/temporary conditions. These temporary situations often generate loads that are outside the original plant design basis. Consequently, separate evaluations are needed to ensure that plant structures, systems and components (SSCs) maintain their integrity and functionality while these temporary loads are active. Also, the temporary structures and components need to be evaluated to ensure their integrity during the temporary duration of use. Three types of approaches are normally adopted either individually or in combination to perform needed evaluations: relax the design allowables, use a more refined analysis model but retain the design basis acceptance criteria, or offset temporary loads by eliminating or reducing part of the design basis loads based on short duration considerations. This paper reviews temporary loading/temporary condition issues and the current industry criteria and procedures proposed in dealing with these issues. Where appropriate, regulatory positions on temporary loads/temporary conditions are discussed.

  17. Safety and Nonsafety Communications and Interactions in International Nuclear Power Plants

    SciTech Connect (OSTI)

    Kisner, Roger A [ORNL; Mullens, James Allen [ORNL; Wilson, Thomas L [ORNL; Wood, Richard Thomas [ORNL; Korsah, Kofi [ORNL; Qualls, A L [ORNL; Muhlheim, Michael David [ORNL; Holcomb, David Eugene [ORNL; Loebl, Andy [ORNL

    2007-08-01T23:59:59.000Z

    Current industry and NRC guidance documents such as IEEE 7-4.3.2, Reg. Guide 1.152, and IEEE 603 do not sufficiently define a level of detail for evaluating interdivisional communications independence. The NRC seeks to establish criteria for safety systems communications that can be uniformly applied in evaluation of a variety of safety system designs. This report focuses strictly on communication issues related to data sent between safety systems and between safety and nonsafety systems. Further, the report does not provide design guidance for communication systems nor present detailed failure modes and effects analysis (FMEA) results for existing designs. This letter report describes communications between safety and nonsafety systems in nuclear power plants outside the United States. A limited study of international nuclear power plants was conducted to ascertain important communication implementations that might have bearing on systems proposed for licensing in the United States. This report provides that following information: 1.communications types and structures used in a representative set of international nuclear power reactors, and 2.communications issues derived from standards and other source documents relevant to safety and nonsafety communications. Topics that are discussed include the following: communication among redundant safety divisions, communications between safety divisions and nonsafety systems, control of safety equipment from a nonsafety workstation, and connection of nonsafety programming, maintenance, and test equipment to redundant safety divisions during operation. Information for this report was obtained through publicly available sources such as published papers and presentations. No proprietary information is represented.

  18. ENVIRONMENTAL PROBLEMS ASSOCIATED WITH DECOMMISSIONING THE CHERNOBYL NUCLEAR POWER PLANT COOLING POND

    SciTech Connect (OSTI)

    Farfan, E.

    2009-09-30T23:59:59.000Z

    Decommissioning of nuclear power plants and other nuclear fuel cycle facilities has been an imperative issue lately. There exist significant experience and generally accepted recommendations on remediation of lands with residual radioactive contamination; however, there are hardly any such recommendations on remediation of cooling ponds that, in most cases, are fairly large water reservoirs. The literature only describes remediation of minor reservoirs containing radioactive silt (a complete closure followed by preservation) or small water reservoirs resulting in reestablishing natural water flows. Problems associated with remediation of river reservoirs resulting in flooding of vast agricultural areas also have been described. In addition, the severity of environmental and economic problems related to the remedial activities is shown to exceed any potential benefits of these activities. One of the large, highly contaminated water reservoirs that require either remediation or closure is Karachay Lake near the MAYAK Production Association in the Chelyabinsk Region of Russia where liquid radioactive waste had been deep well injected for a long period of time. Backfilling of Karachay Lake is currently in progress. It should be noted that secondary environmental problems associated with its closure are considered to be of less importance since sustaining Karachay Lake would have presented a much higher radiological risk. Another well-known highly contaminated water reservoir is the Chernobyl Nuclear Power Plant (ChNPP) Cooling Pond, decommissioning of which is planned for the near future. This study summarizes the environmental problems associated with the ChNPP Cooling Pond decommissioning.

  19. Environmental Problems Associated With Decommissioning The Chernobyl Nuclear Power Plant Cooling Pond

    SciTech Connect (OSTI)

    Farfan, E. B.; Jannik, G. T.; Marra, J. C.; Oskolkov, B. Ya.; Bondarkov, M. D.; Gaschak, S. P.; Maksymenko, A. M.; Maksymenko, V. M.; Martynenko, V. I.

    2009-11-09T23:59:59.000Z

    Decommissioning of nuclear power plants and other nuclear fuel cycle facilities has been an imperative issue lately. There exist significant experience and generally accepted recommendations on remediation of lands with residual radioactive contamination; however, there are hardly any such recommendations on remediation of cooling ponds that, in most cases, are fairly large water reservoirs. The literature only describes remediation of minor reservoirs containing radioactive silt (a complete closure followed by preservation) or small water reservoirs resulting in reestablishing natural water flows. Problems associated with remediation of river reservoirs resulting in flooding of vast agricultural areas also have been described. In addition, the severity of environmental and economic problems related to the remedial activities is shown to exceed any potential benefits of these activities. One of the large, highly contaminated water reservoirs that require either remediation or closure is Karachay Lake near the MAYAK Production Association in the Chelyabinsk Region of Russia where liquid radioactive waste had been deep well injected for a long period of time. Backfilling of Karachay Lake is currently in progress. It should be noted that secondary environmental problems associated with its closure are considered to be of less importance since sustaining Karachay Lake would have presented a much higher radiological risk. Another well-known highly contaminated water reservoir is the Chernobyl Nuclear Power Plant (ChNPP) Cooling Pond, decommissioning of which is planned for the near future. This study summarizes the environmental problems associated with the ChNPP Cooling Pond decommissioning.

  20. Challenges in Determining the Isotopic Mixture for the Fukushima Daiichi Nuclear Power Plant

    SciTech Connect (OSTI)

    Shanks, Arthur [Sandia National Laboratories; Fournier, Sean [Sandia National Laboratories; Shanks, Sonoya [Sandia National Laboratories

    2012-05-01T23:59:59.000Z

    As part of the United States response to the Fukushima Daiichi Nuclear Power Plant emergency, the National Nuclear Security Administration (NNSA) Consequence Management (CM) Teams were activated with elements deploying to Japan. The NNSA CM teams faced the urgent need for information regarding the potential radiological doses that citizens of might experience. This paper discusses the challenges and lessons learned associated with the analysis of field collected samples and gamma spectra in an attempt to determine the isotopic mixture present on the ground around the Plant. There were several interesting and surprising lessons to be learned from the sample analysis portion of the response. The paper discusses several elements of the response that were unique to the event occurring in Japan, as well as several elements that would have occurred in a U.S. nuclear reactor event. Sections of this paper address details of the specific analytical challenges faced during the efforts to analyze samples and try to understand the overall release source term.

  1. A practical approach to risk-based inservice inspection in U.S. nuclear power plants

    SciTech Connect (OSTI)

    Gosselin, S.R. [Electric Power Research Inst., Charlotte, NC (United States); Gamble, R. [Sartrex Corp., Rockville, MD (United States); Dimitrijevic, V.B.; O`Regan, P.J.; Chapman, J.R. [Yankee Atomic Electric Co., Bolton, MS (United States)

    1996-12-01T23:59:59.000Z

    To provide guidelines for practical implementation of risk-based ISI, EPRI sponsored work to develop evaluation procedures and criteria for defining risk-based inservice inspection programs for nuclear power plant piping. These procedures and criteria include efficient means to identify risk significant piping segments, inspection locations, and available inspection techniques. These procedures were applied in a pilot study to assess the feasibility of successfully implementing risk-based inservice inspection programs at nuclear plants. The results from the pilot study indicate that implementation of risk-based inservice inspection programs can reduce the cost and radiation exposure associated with inservice inspection, while maintaining a high level of safety. The list of references provides additional details of these procedures and plant-specific applications. Also, an EPRI technical report has been published to document these procedures. Software has been developed to support and fully document this procedure. Additional development is adding an expert system to the present data base system. The approach compares well to approaches used (or being considered) in other industries and can easily be adapted to these other industries and to address economic and personnel safety in addition to public safety measures.

  2. Recommendations for NRC policy on shift scheduling and overtime at nuclear power plants

    SciTech Connect (OSTI)

    Lewis, P.M.

    1985-07-01T23:59:59.000Z

    This report contains the Pacific Northwest Laboratory's (PNL's) recommendations to the US Nuclear Regulatory Commission (NRC) for an NRC policy on shift scheduling and hours of work (including overtime) for control room operators and other safety-related personnel in nuclear power plants. First, it is recommended that NRC make three additions to its present policy on overtime: (1) limit personnel to 112 hours of work in a 14-day period, 192 hours in 28 days, and 2260 hours in one year; exceeding these limits would require plant manager approval; (2) add a requirement that licensees obtain approval from NRC if plant personnel are expected to exceed 72 hours of work in a 7-day period, 132 hours in 14 days, 228 hours in 28 days, and 2300 hours in one year; and (3) make the policy a requirement, rather than a nonbinding recommendation. Second, it is recommended that licensees be required to obtain NRC approval to adopt a routine 12-hour/day shift schedule. Third, it is recommended that NRC add several nonbinding recommendations concerning routine 8-hour/day schedules. Finally, because additional data can strengthen the basis for future NRC policy on overtime, five methods are suggested for collecting data on overtime and its effects. 44 refs., 10 tabs.

  3. Protection of Nuclear Plants Against Vehicular Bombs Via Full Spectrum Risk Assessment

    SciTech Connect (OSTI)

    Campagna, M. S.; Sawruk, W.

    2003-02-25T23:59:59.000Z

    A more urgent need now exists since 9/11 to protect vital assets at nuclear plants from physical security threats. Any approach to successful defense must result in the best possible risk profile , while also performing this defense against credible threats within the context of limited personnel and materiel resources. Engineered solutions need to be well thought out, and take advantage of each plant's available organic strengths and opportunities. A robust, well trained/equipped highly motivated protective force will help reduce concerns where there are weaknesses making the plant vulnerable to threats. A thorough risk assessment takes into account the proper combination of both deterministic and probabilistic application of resources as a most advantageous approach; this is postulated to be development of integrated protection methods and plans, which blend solid engineering design with the highest caliber of protection forces. By setting a clear and ambitious objective to shield the nuclear assets with this type of dynamic full spectrum defense in depth, the risk of harm-breach or likelihood of any opponent's threat being realized should be reduced to the lowest practicable levels.

  4. An Integrated Scheme for Anomaly Identification and Automatic Control of Nuclear Power Plants Xin Jin, Robert M. Edwards and Asok Ray

    E-Print Network [OSTI]

    Ray, Asok

    An Integrated Scheme for Anomaly Identification and Automatic Control of Nuclear Power Plants Xin.edu INTRODUCTION Nuclear Power Plants (NPPs) are complex systems with many variables that require adjustment Jin, Robert M. Edwards and Asok Ray Department of Mechanical and Nuclear Engineering, The Pennsylvania

  5. Knowledges and abilities catalog for nuclear power plant operators: pressurized water reactors

    SciTech Connect (OSTI)

    Not Available

    1985-07-01T23:59:59.000Z

    This document catalogs roughly 5300 knowledges and abilities of reactor operators and senior reactor operators. It results from a reanalysis of much larger job-task analysis data base compiled by the Institute of Nuclear Power Operations (INPO). Knowledges and abilities are cataloged for 45 major power plant systems and 38 emergency evolutions, grouped according to 11 fundamental safety functions (e.g., reactivity control and reactor coolant system inventory control). With appropriate sampling from this catalog, operator licensing examinations having content validity can be developed. A structured sampling procedure for this catalog is under development by the Nuclear Regulatory Commission (NRC) and will be published as a companion document, ''Examiners' Handbook for Developing Operator Licensing Examinations'' (NUREG-1121). The examinations developed by using the catalog and handbook will cover those topics listed under Title 10, Code of Federal Regulations, Part 55.

  6. A Review of Sensor Calibration Monitoring for Calibration Interval Extension in Nuclear Power Plants

    SciTech Connect (OSTI)

    Coble, Jamie B.; Meyer, Ryan M.; Ramuhalli, Pradeep; Bond, Leonard J.; Hashemian, Hash; Shumaker, Brent; Cummins, Dara

    2012-08-31T23:59:59.000Z

    Currently in the United States, periodic sensor recalibration is required for all safety-related sensors, typically occurring at every refueling outage, and it has emerged as a critical path item for shortening outage duration in some plants. Online monitoring can be employed to identify those sensors that require calibration, allowing for calibration of only those sensors that need it. International application of calibration monitoring, such as at the Sizewell B plant in United Kingdom, has shown that sensors may operate for eight years, or longer, within calibration tolerances. This issue is expected to also be important as the United States looks to the next generation of reactor designs (such as small modular reactors and advanced concepts), given the anticipated longer refueling cycles, proposed advanced sensors, and digital instrumentation and control systems. The U.S. Nuclear Regulatory Commission (NRC) accepted the general concept of online monitoring for sensor calibration monitoring in 2000, but no U.S. plants have been granted the necessary license amendment to apply it. This report presents a state-of-the-art assessment of online calibration monitoring in the nuclear power industry, including sensors, calibration practice, and online monitoring algorithms. This assessment identifies key research needs and gaps that prohibit integration of the NRC-approved online calibration monitoring system in the U.S. nuclear industry. Several needs are identified, including the quantification of uncertainty in online calibration assessment; accurate determination of calibration acceptance criteria and quantification of the effect of acceptance criteria variability on system performance; and assessment of the feasibility of using virtual sensor estimates to replace identified faulty sensors in order to extend operation to the next convenient maintenance opportunity. Understanding the degradation of sensors and the impact of this degradation on signals is key to developing technical basis to support acceptance criteria and set point decisions, particularly for advanced sensors which do not yet have a cumulative history of operating performance.

  7. Formation of hot particles during the Chernobyl nuclear power plant accident

    SciTech Connect (OSTI)

    Kashparov, V.A.; Ivanov, Y.A.; Zvarisch, S.I.; Protsak, V.P.; Khomutinin, Y.V.; Kurepin, A.D.; Pazukhin, E.M. [Ukrainian Inst. of Agricultural Radiology, Chabany (Ukraine)

    1996-05-01T23:59:59.000Z

    The oxidation of irradiated Chernobyl nuclear fuel at 670 to 1,170 K for 3 to 21 h resulted in its destruction into fine particles, the dispersal composition of which is well described by lognormal distribution regularity. The median radius of the formed particles does not depend on the annealing temperature and decreases with the increase of the annealing period from 10 to 3 {micro}m. Proceeding from the dispersal composition and matrix composition of the Chernobyl hot fuel particles, it can be concluded that the oxidation of nuclear fuel was one of the basic mechanisms of hot fuel particle formation during the accident at the Chernobyl nuclear power plant. With oxidation in air and the dispersal of irradiated oxide nuclear fuel at as low as 670 K, ruthenium, located on the granular borders, is released. Ruthenium is oxidized to volatile RuO{sub 4}, sublimated, and condensed on materials of iron. Nickel and stainless steel can be efficiently used at high temperatures (tested to 1,200 K) for radioruthenium adsorption in accidents and for some technological operations. As the temperature of hot fuel particles annealed in inert media increases from 1,270 to 2,270 K, the relative release of radionuclides increases in the following sequence: cesium isotopes; europium isotopes; cerium isotopes; americium isotopes; and ruthenium, plutonium, and curium isotopes.

  8. Estimation of the caesium-137 source term from the Fukushima Daiichi nuclear power plant using a consistent joint assimilation of air concentration and

    E-Print Network [OSTI]

    Boyer, Edmond

    plants in Japan. Diesel backup power sys- tems should have sustained the reactors cooling processEstimation of the caesium-137 source term from the Fukushima Daiichi nuclear power plant using during the accident of the Fukushima Daiichi nuclear power plant in March 2011. In Winiarek et al. (2012b

  9. Radioactive Releases Impact from Kozloduy Nuclear Power Plant, Bulgaria into the Environment

    SciTech Connect (OSTI)

    Genchev, G. T.; Kuleff, I.; Tanev, N. T.; Delistoyanova, E. S.; Guentchev, T.

    2002-02-26T23:59:59.000Z

    The aim of this paper is to present a general overview of the radioactive releases impact generated by Kozloduy Nuclear Power Plant (KNPP), Bulgaria to the environment and public. The liquid releases presented are known as the so called controlled water discharges, that are generated after reprocessing of the inevitable accumulated liquid radioactive waste in the plant operation process. The radionuclides containing in the liquid releases are given in the paper as a result of systematic measuring. Database for radiation doses evaluation on the public around Kozloduy NPP site is developed using IAEA LADTAP computerized program. The computer code LADTAP represents realization of a model that evaluates the public dose as a result of NPP releases under normal operation conditions. The results of this evaluation were the basic licensing document for a new liquid release limit.

  10. Digital control systems in nuclear power plants: Failure information, modeling concepts, and applications. Revision 1

    SciTech Connect (OSTI)

    Galyean, W.J.

    1993-06-23T23:59:59.000Z

    This report briefly describes some current applications of advanced computerized digital display and control systems at US commercial nuclear power plants and presents the results of a literature search that was made to gather information on the reliability of these systems. Both hardware and software reliability were addressed in this review. Only limited failure rate information was found, with the chemical process industry being the primary source of information on hardware failure rates and expert opinion the primary source for software failure rates. Safety-grade digital control systems are typically installed on a functional like-for-like basis, replacing older analog systems without substantially changing interactions with other plant systems. Future work includes performing a limited probabilistic risk assessment of a representative DCS to assess its risk significance.

  11. Digital control systems in nuclear power plants: Failure information, modeling concepts, and applications

    SciTech Connect (OSTI)

    Galyean, W.J.

    1993-06-23T23:59:59.000Z

    This report briefly describes some current applications of advanced computerized digital display and control systems at US commercial nuclear power plants and presents the results of a literature search that was made to gather information on the reliability of these systems. Both hardware and software reliability were addressed in this review. Only limited failure rate information was found, with the chemical process industry being the primary source of information on hardware failure rates and expert opinion the primary source for software failure rates. Safety-grade digital control systems are typically installed on a functional like-for-like basis, replacing older analog systems without substantially changing interactions with other plant systems. Future work includes performing a limited probabilistic risk assessment of a representative DCS to assess its risk significance.

  12. Project plan remove special nuclear material from PFP project plutonium finishing plant

    SciTech Connect (OSTI)

    BARTLETT, W.D.

    1999-05-13T23:59:59.000Z

    This plan presents the overall objectives, description, justification and planning for the Plutonium Finishing Plant (PFP) Remove Special Nuclear Material (SNM) Materials. The intent of this plan is to describe how this project will be managed and integrated with other facility stabilization and deactivation activities. This plan supplements the overall integrated plan presented in the Plutonium Finishing Plant Integrated Project Management Plan (IPMP), HNF-3617, Rev. 0. This project plan is the top-level definitive project management document for PFP Remove SNM Materials project. It specifies the technical, schedule, requirements and the cost baselines to manage the execution of the Remove SNM Materials project. Any deviations to the document must be authorized through the appropriate change control process.

  13. Handbook of human-reliability analysis with emphasis on nuclear power plant applications. Final report

    SciTech Connect (OSTI)

    Swain, A D; Guttmann, H E

    1983-08-01T23:59:59.000Z

    The primary purpose of the Handbook is to present methods, models, and estimated human error probabilities (HEPs) to enable qualified analysts to make quantitative or qualitative assessments of occurrences of human errors in nuclear power plants (NPPs) that affect the availability or operational reliability of engineered safety features and components. The Handbook is intended to provide much of the modeling and information necessary for the performance of human reliability analysis (HRA) as a part of probabilistic risk assessment (PRA) of NPPs. Although not a design guide, a second purpose of the Handbook is to enable the user to recognize error-likely equipment design, plant policies and practices, written procedures, and other human factors problems so that improvements can be considered. The Handbook provides the methodology to identify and quantify the potential for human error in NPP tasks.

  14. Evaluation of aged concrete structures for continued service in nuclear power plants

    SciTech Connect (OSTI)

    Naus, D.J.; Marchbanks, M.F.; Arndt, E.G.

    1988-01-01T23:59:59.000Z

    Results are summarized of a study on concrete component aging and its significance relative to continued service of nuclear power plants (NPPs) beyond the initial period for which they were granted operating licenses. Progress is presented of a second study being conducted to identify and provide acceptance criteria for structural safety issues which the USNRC staff will need to address when applications are submitted for continued service of NPPs. Major activities under this program include: development of a materials property data base, establishment of structural component assessment and repair procedures, and development of a methodology for determination of structural reliability. 19 refs., 5 figs., 3 tabs.

  15. Evaluation of aged concrete structures for continued service in nuclear power plants

    SciTech Connect (OSTI)

    Naus, D.J.; Marchbanks, M.F.; Arndt, E.G.

    1988-01-01T23:59:59.000Z

    Results are summarized of a study on concrete component aging and its significance relative to continued service of nuclear power plants (NPPs) beyond the initial period for which they were granted operating licenses. Progress is presented of a second study being conducted to identify and provide acceptance criteria for structural safety issues which the USNRC staff will need to address when applications are submitted for continued service of NPPs. Major activities under this program include: development of a materials property data base, establishment of structural component assessment and repair procedures, and development of a methodology for determination of structural reliability.

  16. FRAMEWORK AND APPLICATION FOR MODELING CONTROL ROOM CREW PERFORMANCE AT NUCLEAR POWER PLANTS

    SciTech Connect (OSTI)

    Ronald L Boring; David I Gertman; Tuan Q Tran; Brian F Gore

    2008-09-01T23:59:59.000Z

    This paper summarizes an emerging project regarding the utilization of high-fidelity MIDAS simulations for visualizing and modeling control room crew performance at nuclear power plants. The key envisioned uses for MIDAS-based control room simulations are: (i) the estimation of human error associated with advanced control room equipment and configurations, (ii) the investigative determination of contributory cognitive factors for risk significant scenarios involving control room operating crews, and (iii) the certification of reduced staffing levels in advanced control rooms. It is proposed that MIDAS serves as a key component for the effective modeling of cognition, elements of situation awareness, and risk associated with human performance in next generation control rooms.

  17. Preparation of SARs for nonreactor nuclear facilities at the Savannah River Plant

    SciTech Connect (OSTI)

    Durant, W.S.

    1981-01-01T23:59:59.000Z

    Safety Analysis Reports for designated nonreactor nuclear facilities at the Savannah River Plant are prepared in accordance with the DOE Savannah River Manual Chapter 52X1. The accident analysis section is based on the Integrated Risk Assessment Plan, a methodology developed by the Savannah River Laboratory for reprocessing facilities. In general, designated facilities contain radioactive, chemical, or other materials to the extent that a credible accident could have a significant detrimental effect on health and safety. The responsibility for specifying which facilities are designated rests with the manager, Savannah River Operations Office.

  18. Proceedings of the Third International Workshop on the implementation of ALARA at nuclear power plants

    SciTech Connect (OSTI)

    Khan, T.A. [comp.] [Brookhaven National Lab., Upton, NY (United States); Roecklein, A.K. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Regulatory Applications

    1995-03-01T23:59:59.000Z

    This report contains the papers presented and the discussions that took place at the Third International Workshop on ALARA Implementation at Nuclear Power Plants, held in Hauppauge, Long Island, New York from May 8--11, 1994. The purpose of the workshop was to bring together scientists, engineers, health physicists, regulators, managers and other persons who are involved with occupational dose control and ALARA issues. The countries represented were: Canada, Finland, France, Germany, Japan, Korea, Mexico, the Netherlands, Spain, Sweden, the United Kingdom and the United States. The workshop was organized into twelve sessions and three panel discussions. Individual papers have been cataloged separately.

  19. Lead test assembly irradiation and analysis Watts Bar Nuclear Plant, Tennessee and Hanford Site, Richland, Washington

    SciTech Connect (OSTI)

    NONE

    1997-07-01T23:59:59.000Z

    The U.S. Department of Energy (DOE) needs to confirm the viability of using a commercial light water reactor (CLWR) as a potential source for maintaining the nation`s supply of tritium. The Proposed Action discussed in this environmental assessment is a limited scale confirmatory test that would provide DOE with information needed to assess that option. This document contains the environmental assessment results for the Lead test assembly irradiation and analysis for the Watts Bar Nuclear Plant, Tennessee, and the Hanford Site in Richland, Washington.

  20. Procedure for conducting a human-reliability analysis for nuclear power plants. Final report

    SciTech Connect (OSTI)

    Bell, B.J.; Swain, A.D.

    1983-05-01T23:59:59.000Z

    This document describes in detail a procedure to be followed in conducting a human reliability analysis as part of a probabilistic risk assessment when such an analysis is performed according to the methods described in NUREG/CR-1278, Handbook for Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications. An overview of the procedure describing the major elements of a human reliability analysis is presented along with a detailed description of each element and an example of an actual analysis. An appendix consists of some sample human reliability analysis problems for further study.