National Library of Energy BETA

Sample records for water reactor sustainability

  1. Self-Sustaining Thorium Boiling Water Reactors

    E-Print Network [OSTI]

    Ganda, Francesco

    A thorium-fueled water-cooled reactor core design approach that features a radially uniform composition of fuel rods in stationary fuel assembly and is fuel-self-sustaining is described. This core design concept is similar ...

  2. Light Water Reactor Sustainability Program - Non-Destructive...

    Energy Savers [EERE]

    for Determining Remaining Useful Life of Aging Cables in Nuclear Power Plants Light Water Reactor Sustainability Program - Non-Destructive Evaluation R&D Roadmap for...

  3. Light Water Reactor Sustainability Accomplishments Report

    SciTech Connect (OSTI)

    McCarthy, Kathryn A.

    2015-02-01

    Welcome to the 2014 Light Water Reactor Sustainability (LWRS) Program Accomplishments Report, covering research and development highlights from 2014. The LWRS Program is a U.S. Department of Energy research and development program to inform and support the long-term operation of our nation’s commercial nuclear power plants. The research uses the unique facilities and capabilities at the Department of Energy national laboratories in collaboration with industry, academia, and international partners. Extending the operating lifetimes of current plants is essential to supporting our nation’s base load energy infrastructure, as well as reaching the Administration’s goal of reducing greenhouse gas emissions to 80% below 1990 levels by the year 2050. The purpose of the LWRS Program is to provide technical results for plant owners to make informed decisions on long-term operation and subsequent license renewal, reducing the uncertainty, and therefore the risk, associated with those decisions. In January 2013, 104 nuclear power plants operated in 31 states. However, since then, five plants have been shut down (several due to economic reasons), with additional shutdowns under consideration. The LWRS Program aims to minimize the number of plants that are shut down, with R&D that supports long-term operation both directly (via data that is needed for subsequent license renewal), as well indirectly (with models and technology that provide economic benefits). The LWRS Program continues to work closely with the Electric Power Research Institute (EPRI) to ensure that the body of information needed to support SLR decisions and actions is available in a timely manner. This report covers selected highlights from the three research pathways in the LWRS Program: Materials Aging and Degradation, Risk-Informed Safety Margin Characterization, and Advanced Instrumentation, Information, and Control Systems Technologies, as well as a look-ahead at planned activities for 2015. If you have any questions about the information in the report, or about the LWRS Program, please contact me, Richard A. Reister (the Federal Program Manager), or the respective research pathway leader (noted on pages 26 and 27), or visit the LWRS Program website (www.inl.gov/lwrs). The annually updated Integrated Program Plan and Pathway Technical Program Plans are also available for those seeking more detailed technical Information.

  4. Self-Sustaining Thorium Boiling Water Reactors

    SciTech Connect (OSTI)

    Greenspan, Ehud; Gorman, Phillip M.; Bogetic, Sandra; Seifried, Jeffrey E.; Zhang, Guanheng; Varela, Christopher R.; Fratoni, Massimiliano; Vijic, Jasmina J.; Downar, Thomas; Hall, Andrew; Ward, Andrew; Jarrett, Michael; Wysocki, Aaron; Xu, Yunlin; Kazimi, Mujid; Shirvan, Koroush; Mieloszyk, Alexander; Todosow, Michael; Brown, Nicolas; Cheng, Lap

    2015-03-15

    The primary objectives of this project are to: Perform a pre-conceptual design of a core for an alternative to the Hitachi proposed fuel-self- sustaining RBWR-AC, to be referred to as a RBWR-Th. The use of thorium fuel is expected to assure negative void coefficient of reactivity (versus positive of the RBWR-AC) and improve reactor safety; Perform a pre-conceptual design of an alternative core to the Hitachi proposed LWR TRU transmuting RBWR-TB2, to be referred to as the RBWR-TR. In addition to improved safety, use of thorium for the fertile fuel is expected to improve the TRU transmutation effectiveness; Compare the RBWR-Th and RBWR-TR performance against that of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; and, Perform a viability assessment of the thorium-based RBWR design concepts to be identified along with their associated fuel cycle, a technology gap analysis, and a technology development roadmap. A description of the work performed and of the results obtained is provided in this Overview Report and, in more detail, in the Attachments. The major findings of the study are summarized.

  5. Light Water Reactor Sustainability Constellation Pilot Project FY13 Summary Report

    SciTech Connect (OSTI)

    R. Johansen

    2013-09-01

    Summary report for Light Water Reactor Sustainability (LWRS) activities related to the R. E. Ginna and Nine Mile Point Unit 1 for FY13.

  6. Light Water Reactor Sustainability Constellation Pilot Project FY12 Summary Report

    SciTech Connect (OSTI)

    R. Johansen

    2012-09-01

    Summary report for Light Water Reactor Sustainability (LWRS) activities related to the R. E. Ginna and Nine Mile Point Unit 1 for FY12.

  7. Light Water Reactor Sustainability Program Integrated Program Plan

    SciTech Connect (OSTI)

    McCarthy, Kathryn A.; Busby, Jeremy; Hallbert, Bruce; Bragg-Sitton, Shannon; Smith, Curtis; Barnard, Cathy

    2014-04-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

  8. Light Water Reactor Sustainability Program Integrated Program Plan

    SciTech Connect (OSTI)

    Kathryn McCarthy; Jeremy Busby; Bruce Hallbert; Shannon Bragg-Sitton; Curtis Smith; Cathy Barnard

    2013-04-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

  9. Light Water Reactor Sustainability Program Integrated Program Plan

    SciTech Connect (OSTI)

    George Griffith; Robert Youngblood; Jeremy Busby; Bruce Hallbert; Cathy Barnard; Kathryn McCarthy

    2012-01-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline - even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy's Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration's energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program's plans.

  10. Materials Inventory Database for the Light Water Reactor Sustainability Program

    SciTech Connect (OSTI)

    Kazi Ahmed; Shannon M. Bragg-Sitton

    2013-08-01

    Scientific research involves the purchasing, processing, characterization, and fabrication of many sample materials. The history of such materials can become complicated over their lifetime – materials might be cut into pieces or moved to various storage locations, for example. A database with built-in functions to track these kinds of processes facilitates well-organized research. The Material Inventory Database Accounting System (MIDAS) is an easy-to-use tracking and reference system for such items. The Light Water Reactor Sustainability Program (LWRS), which seeks to advance the long-term reliability and productivity of existing nuclear reactors in the United States through multiple research pathways, proposed MIDAS as an efficient way to organize and track all items used in its research. The database software ensures traceability of all items used in research using built-in functions which can emulate actions on tracked items – fabrication, processing, splitting, and more – by performing operations on the data. MIDAS can recover and display the complete history of any item as a simple report. To ensure the database functions suitably for the organization of research, it was developed alongside a specific experiment to test accident tolerant nuclear fuel cladding under the LWRS Advanced Light Water Reactor Nuclear Fuels Pathway. MIDAS kept track of materials used in this experiment from receipt at the laboratory through all processes, test conduct and, ultimately, post-test analysis. By the end of this process, the database proved to be right tool for this program. The database software will help LWRS more efficiently conduct research experiments, from simple characterization tests to in-reactor experiments. Furthermore, MIDAS is a universal tool that any other research team could use to organize their material inventory.

  11. Sustained Recycle in Light Water and Sodium-Cooled Reactors

    SciTech Connect (OSTI)

    Steven J. Piet; Samuel E. Bays; Michael A. Pope; Gilles J. Youinou

    2010-11-01

    From a physics standpoint, it is feasible to sustain recycle of used fuel in either thermal or fast reactors. This paper examines multi-recycle potential performance by considering three recycling approaches and calculating several fuel cycle parameters, including heat, gamma, and neutron emission of fresh fuel; radiotoxicity of waste; and uranium utilization. The first recycle approach is homogeneous mixed oxide (MOX) fuel assemblies in a light water reactor (LWR). The transuranic portion of the MOX was varied among Pu, NpPu, NpPuAm, or all-TRU. (All-TRU means all isotopes through Cf-252.) The Pu case was allowed to go to 10% Pu in fresh fuel, but when the minor actinides were included, the transuranic enrichment was kept below 8% to satisfy the expected void reactivity constraint. The uranium portion of the MOX was enriched uranium. That enrichment was increased (to as much as 6.5%) to keep the fuel critical for a typical LWR irradiation. The second approach uses heterogeneous inert matrix fuel (IMF) assemblies in an LWR - a mix of IMF and traditional UOX pins. The uranium-free IMF fuel pins were Pu, NpPu, NpPuAm, or all-TRU. The UOX pins were limited to 4.95% U-235 enrichment. The number of IMF pins was set so that the amount of TRU in discharged fuel from recycle N (from both IMF and UOX pins) was made into the new IMF pins for recycle N+1. Up to 60 of the 264 pins in a fuel assembly were IMF. The assembly-average TRU content was 1-6%. The third approach uses fast reactor oxide fuel in a sodium-cooled fast reactor with transuranic conversion ratio of 0.50 and 1.00. The transuranic conversion ratio is the production of transuranics divided by destruction of transuranics. The FR at CR=0.50 is similar to the CR for the MOX case. The fast reactor cases had a transuranic content of 33-38%, higher than IMF or MOX.

  12. Light Water Reactor Sustainability Constellation Pilot Project FY11 Summary Report

    SciTech Connect (OSTI)

    R. Johansen

    2011-09-01

    Summary report for Fiscal Year 2011 activities associated with the Constellation Pilot Project. The project is a joint effor between Constellation Nuclear Energy Group (CENG), EPRI, and the DOE Light Water Reactor Sustainability Program. The project utilizes two CENG reactor stations: R.E. Ginna and Nine Point Unit 1. Included in the report are activities associate with reactor internals and concrete containments.

  13. Light Water Reactor Sustainability Program. Digital Architecture Requirements

    SciTech Connect (OSTI)

    Thomas, Kenneth; Oxstrand, Johanna

    2015-03-01

    The Digital Architecture effort is a part of the Department of Energy (DOE) sponsored Light-Water Reactor Sustainability (LWRS) Program conducted at Idaho National Laboratory (INL). The LWRS program is performed in close collaboration with industry research and development (R&D) programs that provides the technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants (NPPs). One of the primary missions of the LWRS program is to help the U.S. nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. Therefore, a major objective of the LWRS program is the development of a seamless digital environment for plant operations and support by integrating information from plant systems with plant processes for nuclear workers through an array of interconnected technologies. In order to get the most benefits of the advanced technology suggested by the different research activities in the LWRS program, the nuclear utilities need a digital architecture in place to support the technology. A digital architecture can be defined as a collection of information technology (IT) capabilities needed to support and integrate a wide-spectrum of real-time digital capabilities for nuclear power plant performance improvements. It is not hard to imagine that many processes within the plant can be largely improved from both a system and human performance perspective by utilizing a plant wide (or near plant wide) wireless network. For example, a plant wide wireless network allows for real time plant status information to easily be accessed in the control room, field workers’ computer-based procedures can be updated based on the real time plant status, and status on ongoing procedures can be incorporated into smart schedules in the outage command center to allow for more accurate planning of critical tasks. The goal of the digital architecture project is to provide a long-term strategy to integrate plant systems, plant processes, and plant workers. This include technologies to improve nuclear worker efficiency and human performance; to offset a range of plant surveillance and testing activities with new on-line monitoring technologies; improve command, control, and collaboration in settings such as outage control centers and work execution centers; and finally to improve operator performance with new operator aid technologies for the control room. The requirements identified through the activities in the Digital Architecture project will be used to estimate the amount of traffic on the network and hence estimating the minimal bandwidth needed.

  14. Overview of the US Department of Energy Light Water Reactor Sustainability Program

    SciTech Connect (OSTI)

    K. A. McCarthy; D. L. Williams; R. Reister

    2012-05-01

    The US Department of Energy Light Water Reactor Sustainability Program is focused on the long-term operation of US commercial power plants. It encompasses two facets of long-term operation: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the nuclear industry that support implementation of performance improvement technologies. An important aspect of the Light Water Reactor Sustainability Program is partnering with industry and the Nuclear Regulatory Commission to support and conduct the long-term research needed to inform major component refurbishment and replacement strategies, performance enhancements, plant license extensions, and age-related regulatory oversight decisions. The Department of Energy research, development, and demonstration role focuses on aging phenomena and issues that require long-term research and/or unique Department of Energy laboratory expertise and facilities and are applicable to all operating reactors. This paper gives an overview of the Department of Energy Light Water Reactor Sustainability Program, including vision, goals, and major deliverables.

  15. Establishment of a Hub for the Light Water Reactor Sustainability Online Monitoring Community

    SciTech Connect (OSTI)

    Nancy J. Lybeck; Magdy S. Tawfik; Binh T. Pham

    2011-08-01

    Implementation of online monitoring and prognostics in existing U.S. nuclear power plants will involve coordinating the efforts of national laboratories, utilities, universities, and private companies. Internet-based collaborative work environments provide necessary communication tools to facilitate interaction between geographically diverse participants. Available technologies were considered, and a collaborative workspace was established at INL as a hub for the light water reactor sustainability online monitoring community.

  16. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

    SciTech Connect (OSTI)

    Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

    2012-09-01

    The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

  17. Light Water Reactor Sustainability Program Status of Silicon Carbide Joining Technology Development

    SciTech Connect (OSTI)

    Shannon M. Bragg-Sitton

    2013-09-01

    Advanced, accident tolerant nuclear fuel systems are currently being investigated for potential application in currently operating light water reactors (LWR) or in reactors that have attained design certification. Evaluation of potential options for accident tolerant nuclear fuel systems point to the potential benefits of silicon carbide (SiC) relative to Zr-based alloys, including increased corrosion resistance, reduced oxidation and heat of oxidation, and reduced hydrogen generation under steam attack (off-normal conditions). If demonstrated to be applicable in the intended LWR environment, SiC could be used in nuclear fuel cladding or other in-core structural components. Achieving a SiC-SiC joint that resists corrosion with hot, flowing water, is stable under irradiation and retains hermeticity is a significant challenge. This report summarizes the current status of SiC-SiC joint development work supported by the Department of Energy Light Water Reactor Sustainability Program. Significant progress has been made toward SiC-SiC joint development for nuclear service, but additional development and testing work (including irradiation testing) is still required to present a candidate joint for use in nuclear fuel cladding.

  18. Report from the Light Water Reactor Sustainability Workshop on On-Line Monitoring Technologies

    SciTech Connect (OSTI)

    Thomas Baldwin; Magdy Tawfik; Leonard Bond

    2010-06-01

    In support of expanding the use of nuclear power, interest is growing in methods of determining the feasibility of longer term operation for the U.S. fleet of nuclear power plants, particularly operation beyond 60 years. To help establish the scientific and technical basis for such longer term operation, the DOE-NE has established a research and development (R&D) objective. This objective seeks to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of current reactors. The Light Water Reactor Sustainability (LWRS) Program, which addresses the needs of this objective, is being developed in collaboration with industry R&D programs to provide the technical foundations for licensing and managing the long-term, safe, and economical operation of nuclear power plants. The LWRS Program focus is on longer-term and higher-risk/reward research that contributes to the national policy objectives of energy and environmental security. In moving to identify priorities and plan activities, the Light Water Reactor Sustainability Workshop on On-Line Monitoring (OLM) Technologies was held June 10–12, 2010, in Seattle, Washington. The workshop was run to enable industry stakeholders and researchers to identify the nuclear industry needs in the areas of future OLM technologies and corresponding technology gaps and research capabilities. It also sought to identify approaches for collaboration that would be able to bridge or fill the technology gaps. This report is the meeting proceedings, documenting the presentations and discussions of the workshop and is intended to serve as a basis for a plan which is under development that will enable the I&C research pathway to achieve its goals. Benefits to the nuclear industry accruing from On Line Monitoring Technology cannot be ignored. Information gathered thus far has contributed significantly to the Department of Energy’s Light Water Reactor Sustainability Program. DOE has shown great interest in supplying necessary support to help this industry to move forward as indicated by the recent workshop conducted in support of this interest. The Light Water Reactor Sustainability Workshop on On-Line Monitoring Technologies provided an opportunity for industry stakeholders and researchers to gather in order to collectively identify the nuclear industry’s needs in the areas of OLM technologies including diagnostics, prognostics, and RUL. Additionally, the workshop provided the opportunity for attendees to pinpoint technology gaps and research capabilities along with the fostering of future collaboration in order to bridge the gaps identified. Attendees concluded that a research and development program is critical to future nuclear operations. Program activities would result in enhancing and modernizing the critical capabilities of instrumentation, information, and control technologies for long-term nuclear asset operation and management. Adopting a comprehensive On Line Monitoring research program intends to: • Develop national capabilities at the university and laboratory level • Create or renew infrastructure needed for long-term research, education, and testing • Support development and testing of needed I&C technologies • Improve understanding of, confidence in, and decisions to employ these new technologies in the nuclear power sector and achieve successful licensing and deployment.

  19. Light Water Reactor Sustainability Program: Analysis of Pressurized Water Reactor Station Blackout caused by external flooding using the RISMC toolkit

    SciTech Connect (OSTI)

    Mandelli, Diego; Smith, Curtis; Prescott, Steven; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua; Kinoshita, Robert

    2014-08-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impacts of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization project aims to provide insights to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This paper focuses on the impacts of power uprate on the safety margin of a boiling water reactor for a flooding induced station black-out event. Analysis is performed by using a combination of thermal-hydraulic codes and a stochastic analysis tool currently under development at the Idaho National Laboratory, i.e. RAVEN. We employed both classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. Results obtained give a detailed investigation of the issues associated with a plant power uprate including the effects of station black-out accident scenarios. We were able to quantify how the timing of specific events was impacted by a higher nominal reactor core power. Such safety insights can provide useful information to the decision makers to perform risk informed margins management.

  20. DOE-NE Light Water Reactor Sustainability Program and EPRI Long Term Operations Program – Joint Research and Development

    Broader source: Energy.gov [DOE]

    Description of Joint DOE and EPRI research and development programs related to reactor sustainability INL/EXT-12-24562

  1. Light Water Reactor Sustainability Program A Reference Plan for Control Room Modernization: Planning and Analysis Phase

    SciTech Connect (OSTI)

    Jacques Hugo; Ronald Boring; Lew Hanes; Kenneth Thomas

    2013-09-01

    The U.S. Department of Energy’s Light Water Reactor Sustainability (LWRS) program is collaborating with a U.S. nuclear utility to bring about a systematic fleet-wide control room modernization. To facilitate this upgrade, a new distributed control system (DCS) is being introduced into the control rooms of these plants. The DCS will upgrade the legacy plant process computer and emergency response facility information system. In addition, the DCS will replace an existing analog turbine control system with a display-based system. With technology upgrades comes the opportunity to improve the overall human-system interaction between the operators and the control room. To optimize operator performance, the LWRS Control Room Modernization research team followed a human-centered approach published by the U.S. Nuclear Regulatory Commission. NUREG-0711, Rev. 3, Human Factors Engineering Program Review Model (O’Hara et al., 2012), prescribes four phases for human factors engineering. This report provides examples of the first phase, Planning and Analysis. The three elements of Planning and Analysis in NUREG-0711 that are most crucial to initiating control room upgrades are: • Operating Experience Review: Identifies opportunities for improvement in the existing system and provides lessons learned from implemented systems. • Function Analysis and Allocation: Identifies which functions at the plant may be optimally handled by the DCS vs. the operators. • Task Analysis: Identifies how tasks might be optimized for the operators. Each of these elements is covered in a separate chapter. Examples are drawn from workshops with reactor operators that were conducted at the LWRS Human System Simulation Laboratory HSSL and at the respective plants. The findings in this report represent generalized accounts of more detailed proprietary reports produced for the utility for each plant. The goal of this LWRS report is to disseminate the technique and provide examples sufficient to serve as a template for other utilities’ projects for control room modernization.

  2. Light Water Reactor Sustainability Research and Development Program Plan -- Fiscal Year 2009–2013

    SciTech Connect (OSTI)

    Idaho National Laboratory

    2009-12-01

    Nuclear power has reliably and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. By the year 2030, domestic demand for electrical energy is expected to grow to levels of 16 to 36% higher than 2007 levels. At the same time, most currently operating nuclear power plants will begin reaching the end of their 60-year operating licenses. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary this year. U.S. regulators have begun considering extended operations of nuclear power plants and the research needed to support long-term operations. The Light Water Reactor Sustainability (LWRS) Research and Development (R&D) Program, developed and sponsored by the Department of Energy, is performed in close collaboration with industry R&D programs. The purpose of the LWRS R&D Program is to provide technical foundations for licensing and managing long-term, safe and economical operation of the current operating nuclear power plants. The LWRS R&D Program vision is captured in the following statements: Existing operating nuclear power plants will continue to safely provide clean and economic electricity well beyond their first license- extension period, significantly contributing to reduction of United States and global carbon emissions, enhancement of national energy security, and protection of the environment. There is a comprehensive technical basis for licensing and managing the long-term, safe, economical operation of nuclear power plants. Sustaining the existing operating U.S. fleet also will improve its international engagement and leadership on nuclear safety and security issues.

  3. Light Water Reactor Sustainability Program Grizzly Year-End Progress Report

    SciTech Connect (OSTI)

    Benjamin Spencer; Yongfeng Zhang; Pritam Chakraborty; S. Bulent Biner; Marie Backman; Brian Wirth; Stephen Novascone; Jason Hales

    2013-09-01

    The Grizzly software application is being developed under the Light Water Reactor Sustainability (LWRS) program to address aging and material degradation issues that could potentially become an obstacle to life extension of nuclear power plants beyond 60 years of operation. Grizzly is based on INL’s MOOSE multiphysics simulation environment, and can simultaneously solve a variety of tightly coupled physics equations, and is thus a very powerful and flexible tool with a wide range of potential applications. Grizzly, the development of which was begun during fiscal year (FY) 2012, is intended to address degradation in a variety of critical structures. The reactor pressure vessel (RPV) was chosen for an initial application of this software. Because it fulfills the critical roles of housing the reactor core and providing a barrier to the release of coolant, the RPV is clearly one of the most safety-critical components of a nuclear power plant. In addition, because of its cost, size and location in the plant, replacement of this component would be prohibitively expensive, so failure of the RPV to meet acceptance criteria would likely result in the shutting down of a nuclear power plant. The current practice used to perform engineering evaluations of the susceptibility of RPVs to fracture is to use the ASME Master Fracture Toughness Curve (ASME Code Case N-631 Section III). This is used in conjunction with empirically based models that describe the evolution of this curve due to embrittlement in terms of a transition temperature shift. These models are based on an extensive database of surveillance coupons that have been irradiated in operating nuclear power plants, but this data is limited to the lifetime of the current reactor fleet. This is an important limitation when considering life extension beyond 60 years. The currently available data cannot be extrapolated with confidence further out in time because there is a potential for additional damage mechanisms (i.e. late blooming phases) to become active later in life beyond the current operational experience. To develop a tool that can eventually serve a role in decision-making, it is clear that research and development must be perfomed at multiple scales. At the engineering scale, a multiphysics analysis code that can capture the thermomechanical response of the RPV under accident conditions, including detailed fracture mechanics evaluations of flaws with arbitrary geometry and orientation, is needed to assess whether the fracture toughness, as defined by the master curve, including the effects of embrittlement, is exceeded. At the atomistic scale, the fundamental mechanisms of degradation need to be understood, including the effects of that degradation on the relevant material properties. In addition, there is a need to better understand the mechanisms leading to the transition from ductile to brittle fracture through improved continuum mechanics modeling at the fracture coupon scale. Work is currently being conducted at all of these levels with the goal of creating a usable engineering tool informed by lower length-scale modeling. This report summarizes progress made in these efforts during FY 2013.

  4. Light Water Reactor Sustainability Program Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study

    SciTech Connect (OSTI)

    Curtis Smith; David Schwieder; Cherie Phelan; Anh Bui; Paul Bayless

    2012-08-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about LWR design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the RISMC Pathway R&D is to support plant decisions for risk-informed margins management with the aim to improve economics, reliability, and sustain safety of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. This report describes the RISMC methodology demonstration where the Advanced Test Reactor (ATR) was used as a test-bed for purposes of determining safety margins. As part of the demonstration, we describe how both the thermal-hydraulics and probabilistic safety calculations are integrated and used to quantify margin management strategies.

  5. Thorium Fuel Options for Sustained Transuranic Burning in Pressurized Water Reactors - 12381

    SciTech Connect (OSTI)

    Rahman, Fariz Abdul; Lee, John C. [University of Michigan, Ann Arbor, MI (United States); Franceschini, Fausto; Wenner, Michael [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

    2012-07-01

    As described in companion papers, Westinghouse is proposing the adoption of a thorium-based fuel cycle to burn the transuranics (TRU) contained in the current Used Nuclear Fuel (UNF) and transition towards a less radio-toxic high level waste. A combination of both light water reactors (LWR) and fast reactors (FR) is envisaged for the task, with the emphasis initially posed on their TRU burning capability and eventually to their self-sufficiency. Given the many technical challenges and development times related to the deployment of TRU burners fast reactors, an interim solution making best use of the current resources to initiate burning the legacy TRU inventory while developing and testing some technologies of later use is desirable. In this perspective, a portion of the LWR fleet can be used to start burning the legacy TRUs using Th-based fuels compatible with the current plants and operational features. This analysis focuses on a typical 4-loop PWR, with 17x17 fuel assembly design and TRUs (or Pu) admixed with Th (similar to U-MOX fuel, but with Th instead of U). Global calculations of the core were represented with unit assembly simulations using the Linear Reactivity Model (LRM). Several assembly configurations have been developed to offer two options that can be attractive during the TRU transmutation campaign: maximization of the TRU transmutation rate and capability for TRU multi-recycling, to extend the option of TRU recycling in LWR until the FR is available. Homogeneous as well as heterogeneous assembly configurations have been developed with various recycling schemes (Pu recycle, TRU recycle, TRU and in-bred U recycle etc.). Oxide as well as nitride fuels have been examined. This enabled an assessment of the potential for burning and multi-recycling TRU in a Th-based fuel PWR to compare against other more typical alternatives (U-MOX and variations thereof). Results will be shown indicating that Th-based PWR fuel is a promising option to multi-recycle and burn TRU in a thermal spectrum, while satisfying top-level operational and safety constraints. Various assembly designs have been proposed to assess the TRU burning potential of Th-based fuel in PWRs. In addition to typical homogeneous loading patterns, heterogeneous configurations exploiting the breeding potential of thorium to enable multiple cycles of TRU irradiation and burning have been devised. The homogeneous assembly design, with all pins featuring TRU in Th, has the benefit of a simple loading pattern and the highest rate of TRU transmutation, but it can be used only for a few cycles due to the rapid rise in the TRU content of the recycled fuel, which challenges reactivity control, safety coefficients and fuel handling. Due to its simple loading pattern, such assembly design can be used as the first step of Th implementation, achieving up to 3 times larger TRU transmutation rate than conventional U-MOX, assuming same fraction of MOX assemblies in the core. As the next step in thorium implementation, heterogeneous assemblies featuring a mixed array of Th-U and Th-U-TRU pins, where the U is in-bred from Th, have been proposed. These designs have the potential to enable burning an external supply of TRU through multiple cycles of irradiation, recovery (via reprocessing) and recycling of the residual actinides at the end of each irradiation cycle. This is achieved thanks to a larger breeding of U from Th in the heterogeneous assemblies, which reduces the TRU supply and thus mitigates the increase in the TRU core inventory for the multi-recycled fuel. While on an individual cycle basis the amount of TRU burned in the heterogeneous assembly is reduced with respect to the homogeneous design, TRU burning rates higher than single-pass U-MOX fuel can still be achieved, with the additional benefits of a multi-cycle transmutation campaign recycling all TRU isotopes. Nitride fuel, due its higher density and U breeding potential, together with its better thermal properties, ideally suits the objectives and constraints of the heterogeneous assemblies. However, signi

  6. Human-In-The-Loop Simulation in Support of Long-Term Sustainability of Light Water Reactors

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Hallbert, Bruce P

    2015-01-01

    Reliable instrumentation, information, and control systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration. The NPP owners and operators realize that this analog technology represents a significant challenge to sustaining the operation of the current fleet of NPPs. Beyond control systems, new technologies are neededmore »to monitor and characterize the effects of aging and degradation in critical areas of key structures, systems, and components. The objective of the efforts sponsored by the U.S. Department of Energy is to develop, demonstrate, and deploy new digital technologies for II&C architectures and provide monitoring capabilities to ensure the continued safe, reliable, and economic operation of the nation’s NPPs.« less

  7. Human-In-The-Loop Simulation in Support of Long-Term Sustainability of Light Water Reactors

    SciTech Connect (OSTI)

    Hallbert, Bruce P

    2015-01-01

    Reliable instrumentation, information, and control systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration. The NPP owners and operators realize that this analog technology represents a significant challenge to sustaining the operation of the current fleet of NPPs. Beyond control systems, new technologies are needed to monitor and characterize the effects of aging and degradation in critical areas of key structures, systems, and components. The objective of the efforts sponsored by the U.S. Department of Energy is to develop, demonstrate, and deploy new digital technologies for II&C architectures and provide monitoring capabilities to ensure the continued safe, reliable, and economic operation of the nation’s NPPs.

  8. Reactor water cleanup system

    DOE Patents [OSTI]

    Gluntz, D.M.; Taft, W.E.

    1994-12-20

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

  9. Reactor water cleanup system

    DOE Patents [OSTI]

    Gluntz, Douglas M. (San Jose, CA); Taft, William E. (Los Gatos, CA)

    1994-01-01

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

  10. Light Water Reactor Sustainability Program FY13 Status Update for EPRI - RISMC Collaboration

    SciTech Connect (OSTI)

    Curtis Smith

    2013-09-01

    The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for risk-informed margins management with the aim to improve economics, reliability, and sustain safety of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced "RISMC toolkit" that enables more accurate representation of NPP safety margin. In order to carry out the R&D needed for the Pathway, the Idaho National Laboratory (INL) is collaborating with the Electric Power Research Institute (EPRI) in order to focus on applications of interest to the U.S. nuclear power industry. This report documents the collaboration activities performed between INL and EPRI during FY2013.

  11. Light Water Reactor Sustainability Program: Computer-based procedure for field activities: results from three evaluations at nuclear power plants

    SciTech Connect (OSTI)

    Oxstrand, Johanna; Bly, Aaron; LeBlanc, Katya

    2014-09-01

    Nearly all activities that involve human interaction with the systems of a nuclear power plant are guided by procedures. The paper-based procedures (PBPs) currently used by industry have a demonstrated history of ensuring safety; however, improving procedure use could yield tremendous savings in increased efficiency and safety. One potential way to improve procedure-based activities is through the use of computer-based procedures (CBPs). Computer-based procedures provide the opportunity to incorporate context driven job aids, such as drawings, photos, just-in-time training, etc into CBP system. One obvious advantage of this capability is reducing the time spent tracking down the applicable documentation. Additionally, human performance tools can be integrated in the CBP system in such way that helps the worker focus on the task rather than the tools. Some tools can be completely incorporated into the CBP system, such as pre-job briefs, placekeeping, correct component verification, and peer checks. Other tools can be partly integrated in a fashion that reduces the time and labor required, such as concurrent and independent verification. Another benefit of CBPs compared to PBPs is dynamic procedure presentation. PBPs are static documents which limits the degree to which the information presented can be tailored to the task and conditions when the procedure is executed. The CBP system could be configured to display only the relevant steps based on operating mode, plant status, and the task at hand. A dynamic presentation of the procedure (also known as context-sensitive procedures) will guide the user down the path of relevant steps based on the current conditions. This feature will reduce the user’s workload and inherently reduce the risk of incorrectly marking a step as not applicable and the risk of incorrectly performing a step that should be marked as not applicable. As part of the Department of Energy’s (DOE) Light Water Reactors Sustainability Program, researchers at Idaho National Laboratory (INL) along with partners from the nuclear industry have been investigating the design requirements for computer-based work instructions (including operations procedures, work orders, maintenance procedures, etc.) to increase efficiency, safety, and cost competitiveness of existing light water reactors.

  12. Light Water Reactor Sustainability Program Support and Modeling for the Boiling Water Reactor Station Black Out Case Study Using RELAP and RAVEN

    SciTech Connect (OSTI)

    Diego Mandelli; Curtis Smith; Thomas Riley; John Schroeder; Cristian Rabiti; Aldrea Alfonsi; Joe Nielsen; Dan Maljovec; Bie Wang; Valerio Pascucci

    2013-09-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated. In order to evaluate the impact of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the impact of power uprate on the safety of a boiled water reactor system. The case study considered is a loss of off-site power followed by the loss of diesel generators, i.e., a station black out (SBO) event. Analysis is performed by using a thermo-hydraulic code, i.e. RELAP-5, and a stochastic analysis tool currently under development at INL, i.e. RAVEN. Starting from the event tree models contained in SAPHIRE, we built the input file for RELAP-5 that models in great detail system dynamics under SBO conditions. We also interfaced RAVEN with RELAP-5 so that it would be possible to run multiple RELAP-5 simulation runs by changing specific keywords of the input file. We both employed classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. We also employed advanced data analysis and visualization tools that helped us to correlate simulation outcome such as maximum core temperature with a set of input uncertain parameters. Results obtained gave a detailed overview of the issues associated to power uprate for a SBO accident scenario. We were able to quantify how timing of safety related events were impacted by a higher reactor core power. Such insights can provide useful material to the decision makers to perform risk-infomed safety margins management.

  13. LIGHT WATER REACTOR SUSTAINABILITY PROGRAM ADVANCED INSTRUMENTATION, INFORMATION, AND CONTROL SYSTEMS TECHNOLOGIES TECHNICAL PROGRAM PLAN FOR 2013

    SciTech Connect (OSTI)

    Hallbert, Bruce; Thomas, Ken

    2014-07-01

    Reliable instrumentation, information, and control (II&C) systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration.

  14. Light Water Reactor Sustainability Program Advanced Instrumentation, Information, and Control Systems Technologies Technical Program Plan for 2013

    SciTech Connect (OSTI)

    Hallbert, Bruce; Thomas, Ken

    2014-09-01

    Reliable instrumentation, information, and control (II&C) systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration.

  15. Reactor Technology | Nuclear Science | ORNL

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

    Reactor Technology Advanced Reactor Concepts Advanced Instrumentation & Controls Light Water Reactor Sustainability Safety and Regulatory Technology Small Modular Reactors Nuclear...

  16. Light Water Reactor Sustainability Program Risk-Informed Safety Margins Characterization (RISMC) PathwayTechnical Program Plan

    SciTech Connect (OSTI)

    Curtis Smith; Cristian Rabiti; Richard Martineau

    2012-11-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). As the current Light Water Reactor (LWR) NPPs age beyond 60 years, there are possibilities for increased frequency of Systems, Structures, and Components (SSCs) degradations or failures that initiate safety-significant events, reduce existing accident mitigation capabilities, or create new failure modes. Plant designers commonly “over-design” portions of NPPs and provide robustness in the form of redundant and diverse engineered safety features to ensure that, even in the case of well-beyond design basis scenarios, public health and safety will be protected with a very high degree of assurance. This form of defense-in-depth is a reasoned response to uncertainties and is often referred to generically as “safety margin.” Historically, specific safety margin provisions have been formulated, primarily based on “engineering judgment.”

  17. Report from the Light Water Reactor Sustainability Workshop on Advanced Instrumentation, Information, and Control Systems and Human-System Interface Technologies

    SciTech Connect (OSTI)

    Bruce P. Hallbert; J. J. Persensky; Carol Smidts; Tunc Aldemir; Joseph Naser

    2009-08-01

    The Light Water Reactor Sustainability (LWRS) Program is a research and development (R&D) program sponsored by the U.S. Department of Energy (DOE). The program is operated in close collaboration with industry R&D programs to provide the technical foundations for licensing and managing the long-term, safe, and economical operation of Nuclear Power Plants that are currently in operation. The LWRS Program focus is on longer-term and higher-risk/reward research that contributes to the national policy objectives of energy and environmental security. Advanced instruments and control (I&C) technologies are needed to support the safe and reliable production of power from nuclear energy systems during sustained periods of operation up to and beyond their expected licensed lifetime. This requires that new capabilities to achieve process control be developed and eventually implemented in existing nuclear assets. It also requires that approaches be developed and proven to achieve sustainability of I&C systems throughout the period of extended operation. The strategic objective of the LWRS Program Advanced Instrumentation, Information, and Control Systems Technology R&D pathway is to establish a technical basis for new technologies needed to achieve safety and reliability of operating nuclear assets and to implement new technologies in nuclear energy systems. This will be achieved by carrying out a program of R&D to develop scientific knowledge in the areas of: • Sensors, diagnostics, and prognostics to support characterization and prediction of the effects of aging and degradation phenomena effects on critical systems, structures, and components (SSCs) • Online monitoring of SSCs and active components, generation of information, and methods to analyze and employ online monitoring information • New methods for visualization, integration, and information use to enhance state awareness and leverage expertise to achieve safer, more readily available electricity generation. As an initial step in accomplishing this effort, the Light Water Reactor Sustainability Workshop on Advanced Instrumentation, Information, and Control Systems and Human-System Interface Technologies was held March 20–21, 2009, in Columbus, Ohio, to enable industry stakeholders and researchers in identification of the nuclear industry’s needs in the areas of future I&C technologies and corresponding technology gaps and research capabilities. Approaches for collaboration to bridge or fill the technology gaps were presented and R&D activities and priorities recommended. This report documents the presentations and discussions of the workshop and is intended to serve as a basis for the plan under development to achieve the goals of the I&C research pathway.

  18. SUSTAINABLE URBAN WATER MANAGEMENT

    E-Print Network [OSTI]

    Das, Suman

    consumption Vehicle production 0.77 0.59 0.79 0.32 4.35 0.44 12.25 2.45 3.85 0.97 (Source: Harto, C; et al% Mining; 1% Decentralized Water Production (LID) Decentralized Energy Production Urban Farming #12;Water Footprint of Agricultural Products #12;`Water for Energy' and `Energy for Water' in US Water for Energy

  19. Sustaining dry surfaces under water

    E-Print Network [OSTI]

    Paul R. Jones; Xiuqing Hao; Eduardo R. Cruz-Chu; Konrad Rykaczewski; Krishanu Nandy; Thomas M. Schutzius; Kripa K. Varanasi; Constantine M. Megaridis; Jens H. Walther; Petros Koumoutsakos; Horacio D. Espinosa; Neelesh A. Patankar

    2014-09-29

    Rough surfaces immersed under water remain practically dry if the liquid-solid contact is on roughness peaks, while the roughness valleys are filled with gas. Mechanisms that prevent water from invading the valleys are well studied. However, to remain practically dry under water, additional mechanisms need consideration. This is because trapped gas (e.g. air) in the roughness valleys can dissolve into the water pool, leading to invasion. Additionally, water vapor can also occupy the roughness valleys of immersed surfaces. If water vapor condenses, that too leads to invasion. These effects have not been investigated, and are critically important to maintain surfaces dry under water. In this work, we identify the critical roughness scale below which it is possible to sustain the vapor phase of water and/or trapped gases in roughness valleys - thus keeping the immersed surface dry. Theoretical predictions are consistent with molecular dynamics simulations and experiments.

  20. Achieving Water-Sustainable Bioenergy Production | Department...

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

    Production Breakout Session 3-A: Growing a Water-Smart Bioeconomy Achieving Water-Sustainable Bioenergy Production May Wu, Principal Environmental System Analyst in the...

  1. First Western Forum on Energy & Water Sustainability

    E-Print Network [OSTI]

    Keller, Arturo A.

    First Western Forum on Energy & Water Sustainability University of California, Santa Barbara John R/Business Support for Comprehensive Energy/Water Program Objective: Support development of water sustainability ­ boils down to economics. Water drives technology and price of energy." "Looking to move to dry or hybrid

  2. Water Markets as a Mechanism for SustainableWater Markets as a Mechanism for Sustainable Reallocation to Meet New Water DemandsReallocation to Meet New Water Demands

    E-Print Network [OSTI]

    Keller, Arturo A.

    Water Markets as a Mechanism for SustainableWater Markets as a Mechanism for SustainableHoover Institution Western Form on Energy and Water Sustainability,Western Form on Energy and Water Sustainability

  3. Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors

    E-Print Network [OSTI]

    Terrani, Kurt Amir

    2010-01-01

    re- actor (PWR) and boiling-water reactor (BWR) designsin integral boiling water super heat reactors. Technical

  4. Stability analysis of supercritical water cooled reactors

    E-Print Network [OSTI]

    Zhao, Jiyun, Ph. D. Massachusetts Institute of Technology

    2005-01-01

    The Supercritical Water-Cooled Reactor (SCWR) is a concept for an advanced reactor that will operate at high pressure (25MPa) and high temperature (500°C average core exit). The high coolant temperature as it leaves the ...

  5. Sustainable Water Management in the Minerals Industry 1 SUSTAINABLE WATER MANAGEMENT IN THE MINERALS

    E-Print Network [OSTI]

    McGuinness, Mark

    is the amount and control of salt content in the used water dams. While there may be several dams at a givenSustainable Water Management in the Minerals Industry 1 SUSTAINABLE WATER MANAGEMENT interest. It arises in the provision of water for Queensland coal mines, where additional water

  6. An Afternoon with Professor Hermanowicz: Exploring Sustainability and Water Filtration

    E-Print Network [OSTI]

    2015-01-01

    different approaches for sustainability and to look at heatwhat constitutes sustainability. BSJ thanks Dr. Hermanowiczdo you measure and define sustainability in the water reuse

  7. First Western Forum on Energy & Water Sustainability

    E-Print Network [OSTI]

    Keller, Arturo A.

    (1987 - 1998) ...Leads to Subsidence in the Central City #12;Colorado River Water Renewable GroundwaterFirst Western Forum on Energy & Water Sustainability March 22, 2007 WATER PLAN: 2000-2050 CITY Provide Service Which Meets Our Customers' Expectations Maximize Use of Renewable Water Supplies Achieve

  8. UC Sustainable Water Systems Policy

    E-Print Network [OSTI]

    Gleeson, Joseph G.

    , or lubricating equipment. Purified Water: Water that is free of impurities such as microorganisms, particulate consumption because it contains objectionable pollution, contamination minerals or infective agents, including water quality standards for human consumption. Reclaimed or Recycled Water: Wastewater treated

  9. Development of Light Water Reactor Fuels with Enhanced Accident...

    Energy Savers [EERE]

    Development of Light Water Reactor Fuels with Enhanced Accident Tolerance - Report to Congress Development of Light Water Reactor Fuels with Enhanced Accident Tolerance - Report to...

  10. Review of light water reactor safety

    SciTech Connect (OSTI)

    Cheng, H.S.

    1980-12-01

    A review of the present status of light water reactor (LWR) safety is presented. The review starts with a brief discussion of the outstanding accident scenarios concerning LWRs. Where possible the areas of present technological uncertainties are stressed. To provide a better perspective of reactor safety, it then reviews the probabilistic assessment of the outstanding LWR accidents considered in the Reactor Safety Study (WASH-1400) and discusses the potential impact of the present technological uncertainties on WASH-1400.

  11. forreading. Sustainability Index for Water Resources

    E-Print Network [OSTI]

    Pasternack, Gregory B.

    -made sustainability indexes are defined for water users in Mexico, the United States, the environment, and for meeting their ecological, environmental and hydrological integrity." Although this concept is still valid, water management impact on a system. Thus, the objective is to look for policies that reduce the adverse impacts of actual

  12. Robustness of RISMC Insights under Alternative Aleatory/Epistemic Uncertainty Classifications: Draft Report under the Risk-Informed Safety Margin Characterization (RISMC) Pathway of the DOE Light Water Reactor Sustainability Program

    SciTech Connect (OSTI)

    Unwin, Stephen D.; Eslinger, Paul W.; Johnson, Kenneth I.

    2012-09-20

    The Risk-Informed Safety Margin Characterization (RISMC) pathway is a set of activities defined under the U.S. Department of Energy (DOE) Light Water Reactor Sustainability Program. The overarching objective of RISMC is to support plant life-extension decision-making by providing a state-of-knowledge characterization of safety margins in key systems, structures, and components (SSCs). A technical challenge at the core of this effort is to establish the conceptual and technical feasibility of analyzing safety margin in a risk-informed way, which, unlike conventionally defined deterministic margin analysis, would be founded on probabilistic characterizations of uncertainty in SSC performance. In the context of probabilistic risk assessment (PRA) technology, there has arisen a general consensus about the distinctive roles of two types of uncertainty: aleatory and epistemic, where the former represents irreducible, random variability inherent in a system, whereas the latter represents a state of knowledge uncertainty on the part of the analyst about the system which is, in principle, reducible through further research. While there is often some ambiguity about how any one contributing uncertainty in an analysis should be classified, there has nevertheless emerged a broad consensus on the meanings of these uncertainty types in the PRA setting. However, while RISMC methodology shares some features with conventional PRA, it will nevertheless be a distinctive methodology set. Therefore, the paradigms for classification of uncertainty in the PRA setting may not fully port to the RISMC environment. Yet the notion of risk-informed margin is based on the characterization of uncertainty, and it is therefore critical to establish a common understanding of uncertainty in the RISMC setting.

  13. Tritium issues in commercial pressurized water reactors

    SciTech Connect (OSTI)

    Jones, G. [Constellation Energy Group, R.E. Ginna Nuclear Power Plant, Ontario, NY (United States)

    2008-07-15

    Tritium has become an important radionuclide in commercial Pressurized Water Reactors because of its mobility and tendency to concentrate in plant systems as tritiated water during the recycling of reactor coolant. Small quantities of tritium are released in routine regulated effluents as liquid water and as water vapor. Tritium has become a focus of attention at commercial nuclear power plants in recent years due to inadvertent, low-level, chronic releases arising from routine maintenance operations and from component failures. Tritium has been observed in groundwater in the vicinity of stations. The nuclear industry has undertaken strong proactive corrective measures to prevent recurrence, and continues to eliminate emission sources through its singular focus on public safety and environmental stewardship. This paper will discuss: production mechanisms for tritium, transport mechanisms from the reactor through plant, systems to the environment, examples of routine effluent releases, offsite doses, basic groundwater transport and geological issues, and recent nuclear industry environmental and legal ramifications. (authors)

  14. NREL: Sustainable NREL - Water Efficiency

    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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration wouldMass map shines lightGeospatial ToolkitSMARTS - Simple ModelCafeteriaVideos of theWater

  15. ORNL). Consortium for Advanced Simulation of Light Water Reactors

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

    Simulation of Light Water Reactors (CASL) was established by the US Department of Energy in 2010 to advance modeling and simulation capabilities for nuclear reactors. CASL's...

  16. Materials Degradation in Light Water Reactors: Life After 60

    Broader source: Energy.gov [DOE]

    Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high temperature water, stress, vibration, and an intense neutron field....

  17. Light Water Reactor Sustainability Program BWR High-Fluence Material Project: Assessment of the Role of High-Fluence on the Efficiency of HWC Mitigation on SCC Crack Growth Rates

    SciTech Connect (OSTI)

    Sebastien Teysseyre

    2014-04-01

    As nuclear power plants age, the increasing neutron fluence experienced by stainless steels components affects the materials resistance to stress corrosion cracking and fracture toughness. The purpose of this report is to identify any new issues that are expected to rise as boiling water reactor power plants reach the end of their initial life and to propose a path forward to study such issues. It has been identified that the efficiency of hydrogen water chemistry mitigation technology may decrease as fluence increases for high-stress intensity factors. This report summarizes the data available to support this hypothesis and describes a program plan to determine the efficiency of hydrogen water chemistry as a function of the stress intensity factor applied and fluence. This program plan includes acquisition of irradiated materials, generation of material via irradiation in a test reactor, and description of the test plan. This plan offers three approaches, each with an estimated timetable and budget.

  18. Light-water reactor accident classification

    SciTech Connect (OSTI)

    Washburn, B.W.

    1980-02-01

    The evolution of existing classifications and definitions of light-water reactor accidents is considered. Licensing practice and licensing trends are examined with respect to terms of art such as Class 8 and Class 9 accidents. Interim definitions, consistent with current licensing practice and the regulations, are proposed for these terms of art.

  19. Cross section generation strategy for high conversion light water reactors

    E-Print Network [OSTI]

    Herman, Bryan R. (Bryan Robert)

    2011-01-01

    High conversion water reactors (HCWR), such as the Resource-renewable Boiling Water Reactor (RBWR), are being designed with axial heterogeneity of alternating fissile and blanket zones to achieve a conversion ratio of ...

  20. Light Water Reactor Fuel Cladding Research and Testing | ornl...

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

    Light Water Reactor Fuel Cladding Research June 01, 2013 Severe Accident Test Station ORNL is the focus point for Light Water Reactor (LWR) fuel cladding research and testing. The...

  1. Containment system for supercritical water oxidation reactor

    DOE Patents [OSTI]

    Chastagner, P.

    1994-07-05

    A system is described for containment of a supercritical water oxidation reactor in the event of a rupture of the reactor. The system includes a containment for housing the reaction vessel and a communicating chamber for holding a volume of coolant, such as water. The coolant is recirculated and sprayed to entrain and cool any reactants that might have escaped from the reaction vessel. Baffles at the entrance to the chamber prevent the sprayed coolant from contacting the reaction vessel. An impact-absorbing layer is positioned between the vessel and the containment to at least partially absorb momentum of any fragments propelled by the rupturing vessel. Remote, quick-disconnecting fittings exterior to the containment, in cooperation with shut-off valves, enable the vessel to be isolated and the system safely taken off-line. Normally-closed orifices throughout the containment and chamber enable decontamination of interior surfaces when necessary. 2 figures.

  2. Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:

    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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirleyEnergyTher i nAandSummary From: JuliaDepartment-8-2008 October5 ofCTIAR.

  3. Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:

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

    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: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankADVANCED MANUFACTURINGEnergy BillsNo. 195 - Oct. 7,DOE HDBK-1113-2008Broadband Plan

  4. Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:

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

    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: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankADVANCED MANUFACTURINGEnergy BillsNo. 195 - Oct. 7,DOE HDBK-1113-2008Broadband PlanInitial Assessment of

  5. Second Forum on Energy & Water Sustainability: Increasing Resource Productivity

    E-Print Network [OSTI]

    Keller, Arturo A.

    in resource efficiency, for energy and water? · What are the co-benefits of implementing these technologiesSecond Forum on Energy & Water Sustainability: Increasing Resource Productivity April 10, 2009 of energy and water sustainability, considering the important linkages between these two resources

  6. Boiling water neutronic reactor incorporating a process inherent safety design

    DOE Patents [OSTI]

    Forsberg, C.W.

    1985-02-19

    A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (nonborated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two water volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

  7. Development of Materials for Supercritical-Water-Cooled Reactor

    Office of Energy Efficiency and Renewable Energy (EERE)

    Supercritical-Water-Cooled Reactor (SCWR) was selected as one of the promising candidates in Generation IV reactors for its prominent advantages; those are the high thermal efficiency, the system...

  8. Boiling water neutronic reactor incorporating a process inherent safety design

    DOE Patents [OSTI]

    Forsberg, Charles W. (Kingston, TN)

    1987-01-01

    A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (non-borated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

  9. Water inventory management in condenser pool of boiling water reactor

    DOE Patents [OSTI]

    Gluntz, Douglas M. (San Jose, CA)

    1996-01-01

    An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs.

  10. Water inventory management in condenser pool of boiling water reactor

    DOE Patents [OSTI]

    Gluntz, D.M.

    1996-03-12

    An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs.

  11. Environmentally assisted cracking in light water reactors

    SciTech Connect (OSTI)

    Chopra, O.K.; Chung, H.M.; Gruber, E.E. [and others

    1996-07-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from April 1995 to December 1995. Topics that have been investigated include fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, EAC of Alloy 600 and 690, and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in simulated LWR environments. Effects of fluoride-ion contamination on susceptibility to intergranular cracking of high- and commercial- purity Type 304 SS specimens from control-tensile tests at 288 degrees Centigrade. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  12. Water resources: sustainable water supply management and basin wide modelling Internationally it has been recognized that the most important challenge to ensuring sustainable

    E-Print Network [OSTI]

    Barthelat, Francois

    Water resources: sustainable water supply management and basin wide modelling Internationally it has been recognized that the most important challenge to ensuring sustainable water use is implementing integrated water resources management (IWRM). It provides the best framework for balancing

  13. Consortium for Advanced Simulation of Light Water Reactors (CASL...

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

    J.C., CASL: Consortium for the Advanced Simulation of Light Water Reactors - A DOE Energy Innovation Hub, ANS MC2015 Joint Internation Conference on Mathematics and Computation...

  14. Consortium for Advanced Simulation of Light Water Reactors (CASL...

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

    Virtual Environment for Scientific Collaboration Posted: April 30, 2013 The Consortium for Advanced Simulation of Light Water Reactors, the Department of Energy's first...

  15. Consortium for Advanced Simulation of Light Water Reactors (CASL...

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

    sensitivities and uncertainties in full-scale multi physics Pressurized Water Reactors (PWR) simulations Research & Development VERA COBRA-TF Dakota Hydra-TH Insilico MAMBA MPACT...

  16. Candidate Materials Evaluation for Supercritical Water-Cooled Reactor

    SciTech Connect (OSTI)

    T. R. Allen and G. S. Was

    2008-12-12

    Final technical report on the corrosion, stress corrosion cracking, and radiation response of candidate materials for the supercritical water-cooled reactor concept.

  17. Commercial Light Water Reactor Tritium Extraction Facility

    SciTech Connect (OSTI)

    McHood, M D

    2000-10-12

    A geotechnical investigation program has been completed for the Commercial Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing, and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  18. State space modeling of reactor core in a pressurized water reactor

    SciTech Connect (OSTI)

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W.; Shamsuddin, Mustaffa; Abdullah, M. Adib

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  19. Driving Water and Wastewater Utilities to More Sustainable Energy Management 

    E-Print Network [OSTI]

    Ferrel, L.; Liner, B.

    2013-01-01

    The Water Environment Federation (WEF) and industry leaders have identified the need for an energy roadmap to guide utilities of all sizes down the road to sustainable energy management through increased renewable energy production, energy...

  20. Analysis of sustainable water supply options for Kuwait

    E-Print Network [OSTI]

    Murtaugh, Katharine A. (Katharine Ann)

    2006-01-01

    This thesis considers several options for improving the sustainability of Kuwait's water supply system. The country currently relies heavily on desalination and brackish groundwater extraction. The options considered for ...

  1. Sustainability Index for Water Resources Planning and Management

    E-Print Network [OSTI]

    Pasternack, Gregory B.

    demonstrating the use of the index. Tailor-made sustainability indexes are defined for water users in Mexico, environmental and hydrological integrity." Although this concept is still valid, water management policies of systems to actual or expected future changes. Vulnerability is the magnitude of an adverse impact

  2. CHARACTERIZATION OF RADIOACTIVITY IN THE REACTOR VESSEL OF THE HEAVY WATER COMPONENT TEST REACTOR

    SciTech Connect (OSTI)

    Vinson, Dennis

    2010-06-01

    The Heavy Water Component Test Reactor (HWCTR) facility is a pressurized heavy water reactor that was used to test candidate fuel designs for heavy water power reactors. The reactor operated at nominal power of 50 MW{sub th}. The reactor coolant loop operated at 1200 psig and 250 C. Two isolated test loop were designed into the reactor to provide special test conditions. Fig. 1 shows a cut-away view of the reactor. The two loops are contained in four inch diameter stainless steel piping. The HWCTR was operated for only a short duration, from March 1962 to December 1964 in order to test the viability of test fuel elements and other reactor components for use in a heavy water power reactor. The reactor achieved 13,882 MWd of total power while testing 36 different fuel assemblies. In the course of operation, HWCTR experienced the cladding failures of 10 separate test fuel assemblies. In each case, the cladding was breached with some release of fuel core material into the isolated test loop, causing fission product and actinide contamination in the main coolant loop and the liquid and boiling test loops. Despite the contribution of the contamination from the failed fuel, the primary source of radioactivity in the HWCTR vessel and internals is the activation products in the thermal shields, and to a lesser degree, activation products in the reactor vessel walls and liner. A detailed facility characterization report of the HWCTR facility was completed in 1996. Many of the inputs and assumptions in the 1996 characterization report were derived from the HWCTR decommissioning plan published in 1975. The current paper provides an updated assessment of the radioisotopic characteristics of the HWCTR vessel and internals to support decommissioning activities on the facility.

  3. Safety of light water reactor fuel with silicon carbide cladding

    E-Print Network [OSTI]

    Lee, Youho

    2013-01-01

    Structural aspects of the performance of light water reactor (LWR) fuel rod with triplex silicon carbide (SiC) cladding - an emerging option to replace the zirconium alloy cladding - are assessed. Its behavior under accident ...

  4. Optimization of hydride fueled pressurized water reactor cores

    E-Print Network [OSTI]

    Shuffler, Carter Alexander

    2004-01-01

    This thesis contributes to the Hydride Fuels Project, a collaborative effort between UC Berkeley and MIT aimed at investigating the potential benefits of hydride fuel use in light water reactors (LWRs). This pursuit involves ...

  5. Process for treating effluent from a supercritical water oxidation reactor

    DOE Patents [OSTI]

    Barnes, C.M.; Shapiro, C.

    1997-11-25

    A method for treating a gaseous effluent from a supercritical water oxidation reactor containing entrained solids is provided comprising the steps of expanding the gas/solids effluent from a first to a second lower pressure at a temperature at which no liquid condenses; separating the solids from the gas effluent; neutralizing the effluent to remove any acid gases; condensing the effluent; and retaining the purified effluent to the supercritical water oxidation reactor. 6 figs.

  6. Stress and Fracture Mechanics Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles

    SciTech Connect (OSTI)

    Yin, Shengjun; Bass, Bennett Richard; Stevens, Gary; Kirk, Mark

    2011-01-01

    This paper describes stress analysis and fracture mechanics work performed to assess boiling water reactor (BWR) and pressurized water reactor (PWR) nozzles located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Various RPV nozzle geometries were investigated: 1. BWR recirculation outlet nozzle; 2. BWR core spray nozzle3 3. PWR inlet nozzle; ; 4. PWR outlet nozzle; and 5. BWR partial penetration instrument nozzle. The above nozzle designs were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-license (EOL) to require evaluation as part of establishing the allowed limits on heatup, cooldown, and hydrotest (leak test) conditions. These nozzles analyzed represent one each of the nozzle types potentially requiring evaluation. The purpose of the analyses performed on these nozzle designs was as follows: To model and understand differences in pressure and thermal stress results using a two-dimensional (2-D) axi-symmetric finite element model (FEM) versus a three-dimensional (3-D) FEM for all nozzle types. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated; To verify the accuracy of a selected linear elastic fracture mechanics (LEFM) hand solution for stress intensity factor for a postulated nozzle corner crack for both thermal and pressure loading for all nozzle types; To assess the significance of attached piping loads on the stresses in the nozzle corner region; and To assess the significance of applying pressure on the crack face with respect to the stress intensity factor for a postulated nozzle corner crack.

  7. Actinide minimization using pressurized water reactors

    E-Print Network [OSTI]

    Visosky, Mark Michael

    2006-01-01

    Transuranic actinides dominate the long-term radiotoxity in spent LWR fuel. In an open fuel cycle, they impose a long-term burden on geologic repositories. Transmuting these materials in reactor systems is one way to ease ...

  8. Rethinking the light water reactor fuel cycle

    E-Print Network [OSTI]

    Shwageraus, Evgeni, 1973-

    2004-01-01

    The once through nuclear fuel cycle adopted by the majority of countries with operating commercial power reactors imposes a number of concerns. The radioactive waste created in the once through nuclear fuel cycle has to ...

  9. Standard practice for evaluation of surveillance capsules from light-water moderated nuclear power reactor vessels

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    Standard practice for evaluation of surveillance capsules from light-water moderated nuclear power reactor vessels

  10. Feasibility study on the thorium fueled boiling water breeder reactor

    SciTech Connect (OSTI)

    PetrusTakaki, N.

    2012-07-01

    The feasibility of (Th,U)O 2 fueled, boiling water breeder reactor based on conventional BWR technology has been studied. In order to determine the potential use of water cooled thorium reactor as a competitive breeder, this study evaluated criticality, breeding and void reactivity coefficient in response to changes made in MFR and fissile enrichments. The result of the study shows that while using light water as moderator, low moderator to fuel volume ratio (MFR=0.5), it was possible to breed fissile fuel in negative void reactivity condition. However the burnup value was lower than the value of the current LWR. On the other hand, heavy water cooled reactor shows relatively wider feasible breeding region, which lead into possibility of designing a core having better neutronic and economic performance than light water with negative void reactivity coefficient. (authors)

  11. Antineutrino monitoring for the Iranian heavy water reactor

    E-Print Network [OSTI]

    Eric Christensen; Patrick Huber; Patrick Jaffke; Thomas Shea

    2014-03-27

    In this note we discuss the potential application of antineutrino monitoring to the Iranian heavy water reactor at Arak, the IR-40, as a non-proliferation measure. We demonstrate that an above ground detector positioned right outside the IR-40 reactor building could meet and in some cases significantly exceed the verification goals identified by IAEA for plutonium production or diversion from declared inventories. In addition to monitoring the reactor during operation, observing antineutrino emissions from long-lived fission products could also allow monitoring the reactor when it is shutdown. Antineutrino monitoring could also be used to distinguish different levels of fuel enrichment. Most importantly, these capabilities would not require a complete reactor operational history and could provide a means to re-establish continuity of knowledge in safeguards conclusions should this become necessary.

  12. "Modeling for effective and sustainable water resources management."

    E-Print Network [OSTI]

    Zhigilei, Leonid V.

    . Current regulations, promulgated at the local and state levels, have focused on impacts of development"Modeling for effective and sustainable water resources management." Teresa Culver Associate Professor tculver@virginia.edu ce.virginia.edu/faculty/culvert.html Dept. of Civil & Environmental

  13. UBC Social Ecological Economic Development Studies (SEEDS) Student Report An Investigation into Sustainable Water Consumption

    E-Print Network [OSTI]

    into Sustainable Water Consumption (Bottled Water versus WaterFillz Units) Jenson Chang, Keqin Chen, Maryam Fallahi of a project/report". #12;An Investigation into Sustainable Water Consumption (Bottled Water versus Water on the environmental, economical, and social impacts of selling over-the-counter bottled waters versus implementing

  14. Noise Decomposition in Boiling Water Reactors with Application to Stability Monitoring

    E-Print Network [OSTI]

    Pázsit, Imre

    Noise Decomposition in Boiling Water Reactors with Application to Stability Monitoring J. Karlsson in boiling water reactor (BWR) noise measure- ments, based on flux factorization techniques (i.e., using reactors4 or flux oscillations in boiling water reactors5,6 ~BWRs!. In these cases the different modes have

  15. Environmental impacts and sustainability of degraded water reuse

    SciTech Connect (OSTI)

    Corwin, D.L.; Bradford, S.A. [USDA ARS, Riverside, CA (United States). US Salin Laboratory

    2008-09-15

    Greater urban demand for finite water resources to meet domestic, agricultural, industrial, and recreational needs; increased frequency of drought resulting from erratic weather; and continued degradation of available water resources from point and nonpoint sources of pollution have focused attention on the reuse of degraded waters as a potential water source. However, short- and long-term detrimental environmental impacts and sustainability of degraded water reuse are not well known or understood. These concerns led to the organization of the 2007 ASA-CSSA-SSSA Symposium entitled Environmental Impacts and Sustainability of Degraded Water Reuse. Out of this symposium came a special collection of 4 review papers and 12 technical research papers focusing on various issues associated with the reuse of agricultural drainage water, well water generated in the production of natural gas from coalbeds, municipal wastewater and biosolids, wastewater from confined animal operations, urban runoff, and food-processing wastewater. Overviews of the papers, gaps in knowledge, and future research directions are presented. The future prognosis of degraded water reuse is promising, provided close attention is paid to managing constituents that pose short- and long-term threats to the environment and the health of humankind.

  16. Sustainable Soil and Water, Adjunct Professor Marcha Johnson Spring 2014 Syllabus (subject to refinement/updating)

    E-Print Network [OSTI]

    Wolberg, George

    Sustainable Soil and Water, Adjunct Professor Marcha Johnson Spring 2014 Syllabus (subject to refinement/updating) 1 Spring 2014 SUS- 7700A: SUSTAINABLE SOIL AND WATER Course: Sustainable Soil and Water water, and groundwater in urban settings; understanding floodplains; treating polluted brownfields

  17. Development of a Heavy Water Detritiation Plant for PIK Reactor

    SciTech Connect (OSTI)

    Alekseev, I.A.; Bondarenko, S.D.; Fedorchenko, O.A.; Konoplev, K.A.; Vasyanina, T.V.; Arkhipov, E.A.; Uborsky, V.V

    2005-07-15

    The research reactor PIK should be supplied with a Detritiation Plant (DP) to remove tritium from heavy water in order to reduce operator radiation dose and tritium emissions. The original design of the reactor PIK Detritiation Plant was completed several years ago. A number of investigations have been made to obtain data for the DP design. Nowadays the design of the DP is being revised on a basis of our investigations. The Combined Electrolysis and Catalytic Exchange (CECE) process will be used at the Detritiation Plant instead of Vapor Phase Catalytic Exchange. The experimental industrial plant for hydrogen isotope separation on the basis of the CECE process is under operation in Petersburg Nuclear Physics Institute. The plant was updated to provide a means for heavy water detritiation. Very high detritiation factors have been achieved in the plant. The use of the CECE process will allow the development of a more compact and less expensive detritiation plant for heavy water reactor PIK.

  18. CAREER: Climate Informed Uncertainty Analyses for Integrated Water Resources Sustainability Principal Investigator: Sankarasubramanian Arumugam

    E-Print Network [OSTI]

    Arumugam, Sankar

    CAREER: Climate Informed Uncertainty Analyses for Integrated Water Resources Sustainability the relative roles of climate variability in modulating seasonal streamflow and water quality variability over forecasts in improving water supply and water quality management and in developing adaptive water management

  19. Advanced Nuclear Technology: Advanced Light Water Reactors Utility Requirements Document Small Modular Reactors Inclusion Summary

    SciTech Connect (OSTI)

    Loflin, Leonard; McRimmon, Beth

    2014-12-18

    This report summarizes a project by EPRI to include requirements for small modular light water reactors (smLWR) into the EPRI Utility Requirements Document (URD) for Advanced Light Water Reactors. The project was jointly funded by EPRI and the U.S. Department of Energy (DOE). The report covers the scope and content of the URD, the process used to revise the URD to include smLWR requirements, a summary of the major changes to the URD to include smLWR, and how to use the URD as revised to achieve value on new plant projects.

  20. Initial Modeling of a Pressurized Water Reactor Completed Using RELAP-7

    Office of Energy Efficiency and Renewable Energy (EERE)

    RELAP-7 is a nuclear reactor system safety analysis code where initial capabilities were demonstrated by simulating a steady-state single-phase pressurized water reactor (PWR) with two parallel loops and multiple reactor core flow channels.

  1. UBC Social Ecological Economic Development Studies (SEEDS) Student Report An Investigation Into Sustainable Water Consumption -

    E-Print Network [OSTI]

    Into Sustainable Water Consumption - Water Bottles versus WaterFillz Units Alireza Tavassoli, Yee Chung Wong, Sina of a project/report". #12;An Investigation Into Sustainable Water Consumption - Water Bottles versus Water, 2010 TA: Craig Hennessey #12;ii Table of Contents Abstract

  2. Biomimicry using Nano-Engineered Enhanced Condensing Surfaces for Sustainable Fresh Water Technology

    E-Print Network [OSTI]

    Al-Beaini, Sara

    2012-01-01

    innovation will enhance a sustainable solution for fresh water produc- tion at a lower costinnovation could o?er a sustainable solution for fresh water production at a lower cost

  3. Development of 1000 MWe Advanced Boiling Water Reactor

    SciTech Connect (OSTI)

    Kazuo Hisajima; Ken Uchida; Keiji Matsumoto; Koichi Kondo; Shigeki Yokoyama; Takuya Miyagawa [Toshiba Corporation (Japan)

    2006-07-01

    1000 MWe Advanced Boiling Water Reactor has only two main steam lines and six reactor internal pumps, whereas 1350 MWe ABWR has four main steam lines and ten reactor internal pumps. In order to confirm how the differences affect hydrodynamic conditions in the dome and lower plenum of the reactor pressure vessel, fluid analyses have been performed. The results indicate that there is not substantial difference between 1000 MWe ABWR and 1350 MWe ABWR. The primary containment vessel of the ABWR consists of the drywell and suppression chamber. The suppression chamber stores water to suppress pressure increase in the primary containment vessel and to be used as the source of water for the emergency core cooling system following a loss-of-coolant accident. Because the reactor pressure vessel of 1000 MWe ABWR is smaller than that of 1350 MWe ABWR, there is room to reduce the size of the primary containment vessel. It has been confirmed feasible to reduce inner diameter of the primary containment vessel from 29 m of 1350 MWe ABWR to 26.5 m. From an economic viewpoint, a shorter outage that results in higher availability of the plant is preferable. In order to achieve 20-day outage that results in 97% of availability, improvement of the systems for removal of decay heat is introduced that enables to stop all the safety-related decay heat removal systems except at the beginning of an outage. (authors)

  4. Rev. 4/1/2009 Second Forum on Energy & Water Sustainability

    E-Print Network [OSTI]

    Keller, Arturo A.

    Rev. 4/1/2009 Second Forum on Energy & Water Sustainability Bren School of Environmental Science (Tibaquira et al.) 9:45 Energy/Water Sustainability and the Electric Power Industry (R. Goldstein et al.) 10 Reductions Related to Energy-Water Sustainability in Sonoma County (G. Keating and D. Pasqualini) 16

  5. Fuel Summary Report: Shippingport Light Water Breeder Reactor - Rev. 2

    SciTech Connect (OSTI)

    Olson, Gail Lynn; Mc Cardell, Richard Keith; Illum, Douglas Brent

    2002-09-01

    The Shippingport Light Water Breeder Reactor (LWBR) was developed by Bettis Atomic Power Laboratory to demonstrate the potential of a water-cooled, thorium oxide fuel cycle breeder reactor. The LWBR core operated from 1977-82 without major incident. The fuel and fuel components suffered minimal damage during operation, and the reactor testing was deemed successful. Extensive destructive and nondestructive postirradiation examinations confirmed that the fuel was in good condition with minimal amounts of cladding deformities and fuel pellet cracks. Fuel was placed in wet storage upon arrival at the Expended Core Facility, then dried and sent to the Idaho Nuclear Technology and Engineering Center for underground dry storage. It is likely that the fuel remains in good condition at its current underground dry storage location at the Idaho Nuclear Technology and Engineering Center. Reports show no indication of damage to the core associated with shipping, loading, or storage.

  6. Assessment of light water reactor accident management programs and experience

    SciTech Connect (OSTI)

    Hammersley, R.J. [Fauske and Associates, Inc., Burr Ridge, IL (United States)

    1992-03-01

    The objective of this report is to provide an assessment of the current light water reactor experience regarding accident management programs and associated technology developments. This assessment for light water reactor (LWR) designs is provided as a resource and reference for the development of accident management capabilities for the production reactors at the Savannah River Site. The specific objectives of this assessment are as follows: 1. Perform a review of the NRC, utility, and industry (NUMARC, EPRI) accident management programs and implementation experience. 2. Provide an assessment of the problems and opportunities in developing an accident management program in conjunction or following the Individual Plant Examination process. 3. Review current NRC, utility, and industry technological developments in the areas of computational tools, severe accident predictive tools, diagnostic aids, and severe accident training and simulation.

  7. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    SciTech Connect (OSTI)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-04-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

  8. The use of reduced-moderation light water reactors for transuranic isotope burning in thorium fuel

    E-Print Network [OSTI]

    Lindley, Benjamin A.

    2015-02-03

    -moderation LWRs can improve TRU transmutation performance, but the VC is still severely limiting for these designs. Reduced-moderation pressurized water reactors (RMPWRs) and boiling water reactors (RBWRs) are considered in this study. Using thorium (Th) instead...

  9. Assessment of innovative fuel designs for high performance light water reactors

    E-Print Network [OSTI]

    Carpenter, David Michael

    2006-01-01

    To increase the power density and maximum allowable fuel burnup in light water reactors, new fuel rod designs are investigated. Such fuel is desirable for improving the economic performance light water reactors loaded with ...

  10. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    SciTech Connect (OSTI)

    Illum, D.B.; Olson, G.L.; McCardell, R.K.

    1999-01-01

    The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.

  11. Upper internals arrangement for a pressurized water reactor

    DOE Patents [OSTI]

    Singleton, Norman R; Altman, David A; Yu, Ching; Rex, James A; Forsyth, David R

    2013-07-09

    In a pressurized water reactor with all of the in-core instrumentation gaining access to the core through the reactor head, each fuel assembly in which the instrumentation is introduced is aligned with an upper internals instrumentation guide-way. In the elevations above the upper internals upper support assembly, the instrumentation is protected and aligned by upper mounted instrumentation columns that are part of the instrumentation guide-way and extend from the upper support assembly towards the reactor head in hue with a corresponding head penetration. The upper mounted instrumentation columns are supported laterally at one end by an upper guide tube and at the other end by the upper support plate.

  12. Light Water Reactor Sustainability Program - Integrated Program Plan |

    Energy Savers [EERE]

    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: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on DeliciousMathematics AndBeryllium Disease | Department of0 Inspection BEFORE9 - Energy andLife Events Life

  13. Light Water Reactor Sustainability Technical Documents | Department of

    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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirleyEnergy A plug-inPPLforLDRDEnergyTurbine bladesJuneLeucadiaDepartment of

  14. Light Water Reactor Sustainability Technical Documents | Department of

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

    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: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankADVANCED MANUFACTURING OFFICE INDUSTRIALU.S.Leadership on Clean EnergysR&D Roadmap

  15. Light Water Reactor Sustainability Nondestructive Evaluation for Concrete

    Energy Savers [EERE]

    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: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on DeliciousMathematicsEnergyInterested Parties - WAPA PublicLED ADOPTIONto CommercializationLight My

  16. Light Water Reactor Sustainability Program - Integrated Program Plan |

    Energy Savers [EERE]

    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: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on DeliciousMathematicsEnergyInterested Parties - WAPA PublicLED ADOPTIONto CommercializationLight MyDepartment of

  17. Light Water Reactor Sustainability Program - Non-Destructive Evaluation

    Energy Savers [EERE]

    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: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on DeliciousMathematicsEnergyInterested Parties - WAPA PublicLED ADOPTIONto CommercializationLight MyDepartment

  18. Light Water Reactor Sustainability Program: Integrated Program Plan |

    Energy Savers [EERE]

    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: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on DeliciousMathematicsEnergyInterested Parties - WAPA PublicLED ADOPTIONto CommercializationLight

  19. Light Water Reactor Sustainability Program: Materials Aging and Degradation

    Energy Savers [EERE]

    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: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on DeliciousMathematicsEnergyInterested Parties - WAPA PublicLED ADOPTIONto CommercializationLightTechnical Program

  20. Light Water Reactor Sustainability (LWRS) Program | Department of Energy

    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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration would likeUniverseIMPACTThousand CubicResourcelogo and-E C H N13, 2009LienertProducts,Nuclear

  1. Accident Performance of Light Water Reactor Cladding Materials

    SciTech Connect (OSTI)

    Nelson, Andrew T.

    2012-07-24

    During a loss of coolant accident as experienced at Fukushima, inadequate cooling of the reactor core forces component temperatures ever higher where they must withstand aggressive chemical environments. Conventional zirconium cladding alloys will readily oxidize in the presence of water vapor at elevated temperatures, rapidly degrading and likely failing. A cladding breach removes the critical barrier between actinides and fission products and the coolant, greatly increasing the probability of the release of radioactivity in the event of a containment failure. These factors have driven renewed international interest in both study and improvement of the materials used in commercial light water reactors. Characterization of a candidate cladding alloy or oxidation mitigation technique requires understanding of both the oxidation kinetics and hydrogen production as a function of temperature and atmosphere conditions. Researchers in the MST division supported by the DOE-NE Fuel Cycle Research and Development program are working to evaluate and quantify these parameters across a wide range of proposed cladding materials. The primary instrument employed is a simultaneous thermal analyzer (STA) equipped with a specialized water vapor furnace capable of maintaining temperatures above 1200 C in a range of atmospheres and water vapor contents. The STA utilizes thermogravimetric analysis and a coupled mass spectrometer to measure in situ oxidation and hydrogen production of candidate materials. This capability is unprecedented in study of materials under consideration for reactor cladding use, and is currently being expanded to investigate proposed coating techniques as well as the effect of coating defects on corrosion resistance.

  2. Pebble Bed Boiling Water Reactor Concept With Superheated Steam

    SciTech Connect (OSTI)

    Tsiklauri, G.; Newman, D.; Meriwether, G.; Korolev, V. [Pacific Northwest National Laboratory, P.O. Box 999 Richland, WA 99352 (United States)

    2002-07-01

    An Advanced Nuclear Reactor concept is presented which extends Boiling Water Reactor technology with micro-fuel elements (MFE) and produces superheated steam. A nuclear plant with MFE is highly efficient and safe, due to ceramic-clad nuclear fuel. Water is used as both moderator and coolant. The fuel consists of spheres of about 1.5 mm diameter of UO{sub 2} with several external coatings of different carbonaceous materials. The outer coating of the particles is SiC, manufactured with chemical vapor disposition (CVD) technology. Endurance of the integrity of the SiC coating in water, air and steam has been demonstrated experimentally in Germany, Russia and Japan. This paper describes a result of a preliminary design and analysis of 3750 MWt (1500 MWe) plant with standard pressure of 16 MPa, which is widely achieved in the vessel of pressurized-water type reactors. The superheated steam outlet temperature of 550 deg. C elevates the steam cycle to high thermal efficiency of 42%. (authors)

  3. Multi-Applications Small Light Water Reactor - NERI Final Report

    SciTech Connect (OSTI)

    S. Michale Modro; James E. Fisher; Kevan D. Weaver; Jose N. Reyes, Jr.; John T. Groome; Pierre Babka; Thomas M. Carlson

    2003-12-01

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle.

  4. Analysis of scrams and forced outages at boiling water reactors

    SciTech Connect (OSTI)

    Earle, R. T.; Sullivan, W. P.; Miller, K. R.; Schwegman, W. J.

    1980-07-01

    This report documents the results of a study of scrams and forced outages at General Electric Boiling Water Reactors (BWRs) operating in the United States. This study was conducted for Sandia Laboratories under a Light Water Reactor Safety Program which it manages for the United States Department of Energy. Operating plant data were used to identify the causes of scrams and forced outages. Causes of scrams and forced outages have been summarized as a function of operating plant and plant age and also ranked according to the number of events per year, outage time per year, and outage time per event. From this ranking, identified potential improvement opportunities were evaluated to determine the associated benefits and impact on plant availability.

  5. Stress corrosion cracking and crack tip characterization of Alloy X-750 in light water reactor environments

    E-Print Network [OSTI]

    Gibbs, Jonathan Paul

    2011-01-01

    Stress corrosion cracking (SCC) susceptibility of Inconel Alloy X-750 in the HTH condition has been evaluated in high purity water at 93 and 288°C under Boiling Water Reactor Normal Water Chemistry (NWC) and Hydrogen Water ...

  6. Stress Corrosion Cracking and Crack Tip Characterization of Alloy X-750 in Light Water Reactor Environments

    E-Print Network [OSTI]

    Gibbs, Jonathan Paul

    Stress corrosion cracking (SCC) susceptibility of Inconel Alloy X-750 in the HTH condition has been evaluated in high purity water at 93 and 288°C under Boiling Water Reactor Normal Water Chemistry (NWC) and Hydrogen Water ...

  7. Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors

    E-Print Network [OSTI]

    Terrani, Kurt Amir

    2010-01-01

    Fundamental aspects of nuclear reactor fuel elements.Unlike permanent nuclear reactor core components, nuclearof the first nuclear reactors, commercial nuclear fuel still

  8. Sustainable Water Practices at Pomona's Parks: Improving Irrigation Use and Stormwater Runoff Retention

    E-Print Network [OSTI]

    Young, Terence

    Sustainable Water Practices at Pomona's Parks: Improving Irrigation Use and Stormwater Runoff;1 Abstract The need for water sustainability has emerged as a response to urbanization and the depletion of water sources. Urban green spaces, such as the city parks of Pomona, are at the forefront for the need

  9. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    SciTech Connect (OSTI)

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  10. Pressurized water nuclear reactor system with hot leg vortex mitigator

    DOE Patents [OSTI]

    Lau, Louis K. S. (Monroeville, PA)

    1990-01-01

    A pressurized water nuclear reactor system includes a vortex mitigator in the form of a cylindrical conduit between the hot leg conduit and a first section of residual heat removal conduit, which conduit leads to a pump and a second section of residual heat removal conduit leading back to the reactor pressure vessel. The cylindrical conduit is of such a size that where the hot leg has an inner diameter D.sub.1, the first section has an inner diameter D.sub.2, and the cylindrical conduit or step nozzle has a length L and an inner diameter of D.sub.3 ; D.sub.3 /D.sub.1 is at least 0.55, D.sub.2 is at least 1.9, and L/D.sub.3 is at least 1.44, whereby cavitation of the pump by a vortex formed in the hot leg is prevented.

  11. Evaluation of the economic simplified boiling water reactor human reliability analysis using the SHARP framework

    E-Print Network [OSTI]

    Dawson, Phillip Eng

    2007-01-01

    General Electric plans to complete a design certification document for the Economic Simplified Boiling Water Reactor to have the new reactor design certified by the United States Nuclear Regulatory Commission. As part of ...

  12. Pressurized water reactor in-core nuclear fuel management by tabu search

    E-Print Network [OSTI]

    Hill, Natasha J.; Parks, Geoffrey T.

    2014-08-24

    Optimization of the arrangement of fuel assemblies and burnable poisons when reloading pressurized water reactors has, in the past, been performed with many di erent algorithms in an attempt to make reactors more economic and fuel effi cient...

  13. The impact of passive safety systems on desirability of advanced light water reactors

    E-Print Network [OSTI]

    Eul, Ryan C

    2006-01-01

    This work investigates whether the advanced light water reactor designs with passive safety systems are more desirable than advanced reactor designs with active safety systems from the point of view of uncertainty in the ...

  14. Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

    SciTech Connect (OSTI)

    Not Available

    1993-05-13

    Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

  15. Enabling Mass-Scale Financing for Federal Energy, Water, and Sustainability Projects

    Broader source: Energy.gov [DOE]

    Presentation covers Enabling Mass-Scale Financing for Federal Energy, Water, and Sustainability projects and is given at the Spring 2011 Federal Utility Partnership Working Group (FUPWG) meeting.

  16. UBC Social Ecological Economic Development Studies (SEEDS) Student Report Sustainable Water Consumption

    E-Print Network [OSTI]

    UBC Social Ecological Economic Development Studies (SEEDS) Student Report Sustainable Water, UBC Sustainability and affiliated organizations have made significant gains in developing resources or the SEEDS Coordinator about the current status of the subject matter of a project/report". #12;Sustainable

  17. The Purdue Water Community Strategic Plan for 2011-2016 Interdisciplinary solutions for sustainable global and regional water resource

    E-Print Network [OSTI]

    The Purdue Water Community Strategic Plan for 2011-2016 Interdisciplinary solutions for sustainable global and regional water resource Vision The Purdue Water Community (PWC) is a vibrant, interconnected information focused on solving global and regional water challenges. Mission The Purdue Water Community

  18. Minor actinide transmutation in thorium and uranium matrices in heavy water moderated reactors

    SciTech Connect (OSTI)

    Bhatti, Zaki; Hyland, B.; Edwards, G.W.R. [Atomic Energy of Canada Limited, Chalk River Laboratories, 1 Plant Road, Chalk River, Ontario, K0J 1J0 (Canada)

    2013-07-01

    The irradiation of Th{sup 232} breeds fewer of the problematic minor actinides (Np, Am, Cm) than the irradiation of U{sup 238}. This characteristic makes thorium an attractive potential matrix for the transmutation of these minor actinides, as these species can be transmuted without the creation of new actinides as is the case with a uranium fuel matrix. Minor actinides are the main contributors to long term decay heat and radiotoxicity of spent fuel, so reducing their concentration can greatly increase the capacity of a long term deep geological repository. Mixing minor actinides with thorium, three times more common in the Earth's crust than natural uranium, has the additional advantage of improving the sustainability of the fuel cycle. In this work, lattice cell calculations have been performed to determine the results of transmuting minor actinides from light water reactor spent fuel in a thorium matrix. 15-year-cooled group-extracted transuranic elements (Np, Pu, Am, Cm) from light water reactor (LWR) spent fuel were used as the fissile component in a thorium-based fuel in a heavy water moderated reactor (HWR). The minor actinide (MA) transmutation rates, spent fuel activity, decay heat and radiotoxicity, are compared with those obtained when the MA were mixed instead with natural uranium and taken to the same burnup. Each bundle contained a central pin containing a burnable neutron absorber whose initial concentration was adjusted to have the same reactivity response (in units of the delayed neutron fraction ?) for coolant voiding as standard NU fuel. (authors)

  19. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    SciTech Connect (OSTI)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

    The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

  20. Camera Inspection Arm for Boiling Water Reactors - 13330

    SciTech Connect (OSTI)

    Martin, Scott; Rood, Marc

    2013-07-01

    Boiling Water Reactor (BWR) outage maintenance tasks can be time-consuming and hazardous. Reactor facilities are continuously looking for quicker, safer, and more effective methods of performing routine inspection during these outages. In 2011, S.A. Technology (SAT) was approached by Energy Northwest to provide a remote system capable of increasing efficiencies related to Reactor Pressure Vessel (RPV) internal inspection activities. The specific intent of the system discussed was to inspect recirculation jet pumps in a manner that did not require manual tooling, and could be performed independently of other ongoing inspection activities. In 2012, SAT developed a compact, remote, camera inspection arm to create a safer, more efficient outage environment. This arm incorporates a compact and lightweight design along with the innovative use of bi-stable composite tubes to provide a six-degree of freedom inspection tool capable of reducing dose uptake, reducing crew size, and reducing the overall critical path for jet pump inspections. The prototype camera inspection arm unit is scheduled for final testing in early 2013 in preparation for the Columbia Generating Station refueling outage in the spring of 2013. (authors)

  1. Supercritical Water Reactor Cycle for Medium Power Applications

    SciTech Connect (OSTI)

    BD Middleton; J Buongiorno

    2007-04-25

    Scoping studies for a power conversion system based on a direct-cycle supercritical water reactor have been conducted. The electric power range of interest is 5-30 MWe with a design point of 20 MWe. The overall design objective is to develop a system that has minimized physical size and performs satisfactorily over a broad range of operating conditions. The design constraints are as follows: Net cycle thermal efficiency {ge}20%; Steam turbine outlet quality {ge}90%; and Pumping power {le}2500 kW (at nominal conditions). Three basic cycle configurations were analyzed. Listed in order of increased plant complexity, they are: (1) Simple supercritical Rankine cycle; (2) All-supercritical Brayton cycle; and (3) Supercritical Rankine cycle with feedwater preheating. The sensitivity of these three configurations to various parameters, such as reactor exit temperature, reactor pressure, condenser pressure, etc., was assessed. The Thermoflex software package was used for this task. The results are as follows: (a) The simple supercritical Rankine cycle offers the greatest hardware simplification, but its high reactor temperature rise and reactor outlet temperature may pose serious problems from the viewpoint of thermal stresses, stability and materials in the core. (b) The all-supercritical Brayton cycle is not a contender, due to its poor thermal efficiency. (c) The supercritical Rankine cycle with feedwater preheating affords acceptable thermal efficiency with lower reactor temperature rise and outlet temperature. (d) The use of a moisture separator improves the performance of the supercritical Rankine cycle with feedwater preheating and allows for a further reduction of the reactor outlet temperature, thus it was selected for the next step. Preliminary engineering design of the supercritical Rankine cycle with feedwater preheating and moisture separation was performed. All major components including the turbine, feedwater heater, feedwater pump, condenser, condenser pump and pipes were modeled with realistic assumptions using the PEACE module of Thermoflex. A three-dimensional layout of the plant was also generated with the SolidEdge software. The results of the engineering design are as follows: (i) The cycle achieves a net thermal efficiency of 24.13% with 350/460 C reactor inlet/outlet temperatures, {approx}250 bar reactor pressure and 0.75 bar condenser pressure. The steam quality at the turbine outlet is 90% and the total electric consumption of the pumps is about 2500 kWe at nominal conditions. (ii) The overall size of the plant is attractively compact and can be further reduced if a printed-circuit-heat-exchanger (vs shell-and-tube) design is used for the feedwater heater, which is currently the largest component by far. Finally, an analysis of the plant performance at off-nominal conditions has revealed good robustness of the design in handling large changes of thermal power and seawater temperature.

  2. An Integrated Water Treatment Technology Solution for Sustainable Water Resource Management in the Marcellus Shale

    SciTech Connect (OSTI)

    Matthew Bruff; Ned Godshall; Karen Evans

    2011-04-30

    This Final Scientific/ Technical Report submitted with respect to Project DE-FE0000833 titled 'An Integrated Water Treatment Technology Solution for Sustainable Water Resource Management in the Marcellus Shale' in support of final reporting requirements. This final report contains a compilation of previous reports with the most current data in order to produce one final complete document. The goal of this research was to provide an integrated approach aimed at addressing the increasing water resource challenges between natural gas production and other water stakeholders in shale gas basins. The objective was to demonstrate that the AltelaRain{reg_sign} technology could be successfully deployed in the Marcellus Shale Basin to treat frac flow-back water. That objective has been successfully met.

  3. Consortium for Advanced Simulation of Light Water Reactors (CASL...

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

    in AP1000 reactor core Test run signals emergence of the next generation in nuclear power reactor analysis tools OAK RIDGE, Tenn., Feb. 18, 2014 - Scientists and engineers...

  4. DOE/NNSA perspective safeguard by design: GEN III/III+ light water reactors and beyond

    SciTech Connect (OSTI)

    Pan, Paul Y

    2010-12-10

    An overview of key issues relevant to safeguards by design (SBD) for GEN III/IV nuclear reactors is provided. Lessons learned from construction of typical GEN III+ water reactors with respect to SBD are highlighted. Details of SBD for safeguards guidance development for GEN III/III+ light water reactors are developed and reported. This paper also identifies technical challenges to extend SBD including proliferation resistance methodologies to other GEN III/III+ reactors (except HWRs) and GEN IV reactors because of their immaturity in designs.

  5. Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor

    SciTech Connect (OSTI)

    Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag; Gilles Youinou; Brian Boer

    2012-04-01

    Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU)-only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.

  6. "Some, for all, forever" Managing Water for Sustainable Development in South Africa

    E-Print Network [OSTI]

    Mauzerall, Denise

    "Some, for all, forever" Managing Water for Sustainable Development in South Africa Priscilla" Integrated Water Resources Management in South Africa" a. Natural Conditions and Hydrology b. Pollution c. Population and a Legacy of Inequality d. New Legislation: The National Water Act of 1998 and the Water

  7. Technologies for Upgrading Light Water Reactor Outlet Temperature

    SciTech Connect (OSTI)

    Daniel S. Wendt; Piyush Sabharwall; Vivek Utgikar

    2013-07-01

    Nuclear energy could potentially be utilized in hybrid energy systems to produce synthetic fuels and feedstocks from indigenous carbon sources such as coal and biomass. First generation nuclear hybrid energy system (NHES) technology will most likely be based on conventional light water reactors (LWRs). However, these LWRs provide thermal energy at temperatures of approximately 300°C, while the desired temperatures for many chemical processes are much higher. In order to realize the benefits of nuclear hybrid energy systems with the current LWR reactor fleets, selection and development of a complimentary temperature upgrading technology is necessary. This paper provides an initial assessment of technologies that may be well suited toward LWR outlet temperature upgrading for powering elevated temperature industrial and chemical processes during periods of off-peak power demand. Chemical heat transformers (CHTs) are a technology with the potential to meet LWR temperature upgrading requirements for NHESs. CHTs utilize chemical heat of reaction to change the temperature at which selected heat sources supply or consume thermal energy. CHTs could directly utilize LWR heat output without intermediate mechanical or electrical power conversion operations and the associated thermodynamic losses. CHT thermal characteristics are determined by selection of the chemical working pair and operating conditions. This paper discusses the chemical working pairs applicable to LWR outlet temperature upgrading and the CHT operating conditions required for providing process heat in NHES applications.

  8. Effect of heat recirculation on the self-sustained catalytic combustion of propane/air mixtures in a quartz reactor

    SciTech Connect (OSTI)

    Scarpa, A. [Dipartimento di Ingegneria Chimica, Universita degli Studi di Napoli ''Federico II'', P.le V. Tecchio 80, 80125 Naples (Italy); Department of Chemical Engineering, Center for Catalytic Science and Technology (CCST), and Center for Composite Materials (CCM), University of Delaware, 150 Academy Street, Newark, DE 19716 (United States); Pirone, R. [Istituto di Ricerche sulla Combustione-CNR, P.le V. Tecchio 80, 80125 Naples (Italy); Russo, G. [Dipartimento di Ingegneria Chimica, Universita degli Studi di Napoli ''Federico II'', P.le V. Tecchio 80, 80125 Naples (Italy); Vlachos, D.G. [Department of Chemical Engineering, Center for Catalytic Science and Technology (CCST), and Center for Composite Materials (CCM), University of Delaware, 150 Academy Street, Newark, DE 19716 (United States)

    2009-05-15

    The self-sustained catalytic combustion of propane is experimentally studied in a two-pass, quartz heat-recirculation reactor (HRR) and compared to that in a no (heat) recirculation reactor (NRR). Structured monolithic reactors with Pt/{gamma}-Al{sub 2}O{sub 3}, LaMnO{sub 3}/{gamma}-Al{sub 2}O{sub 3}, and Pt doped perovskite catalysts have been compared in the HRR and NRR configurations. Heat recirculation enhances combustion stability, by widening the operating window of self-sustained operation, and changes the mode of stability loss from blowout to extinction. It is found that thermal shields (upstream and downstream of the monolith) play no role in the stability of a HRR but increase the stability of a NRR. The stability of a HRR follows this trend: Pt/{gamma}-Al{sub 2}O{sub 3} > doped perovskite > LaMnO{sub 3}/{gamma}-Al{sub 2}O{sub 3}. Finally, a higher cell density monolith enlarges the operating window of self-sustained combustion, and allows further increase of the power density of the process. (author)

  9. Commercial Light Water Reactor Tritium Extraction Facility Geotechnical Summary Report

    SciTech Connect (OSTI)

    Lewis, M.R.

    2000-01-11

    A geotechnical investigation program has been completed for the Circulating Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  10. Technology Implementation Plan. Fully Ceramic Microencapsulated Fuel for Commercial Light Water Reactor Application

    SciTech Connect (OSTI)

    Snead, Lance Lewis; Terrani, Kurt A.; Powers, Jeffrey J.; Worrall, Andrew; Robb, Kevin R.; Snead, Mary A.

    2015-04-01

    This report is an overview of the implementation plan for ORNL's fully ceramic microencapsulated (FCM) light water reactor fuel. The fully ceramic microencapsulated fuel consists of tristructural isotropic (TRISO) particles embedded inside a fully dense SiC matrix and is intended for utilization in commercial light water reactor application.

  11. Water and Energy Sustainability: A Balance of Government Action and Industry Innovation

    SciTech Connect (OSTI)

    Ben Grunewald

    2009-12-31

    By completing the tasks and subtasks of the project, the Ground Water Protection Council (GWPC) through its state regulatory agency members and oil and gas industry partners, will bring attention to water quality and quantity issues and make progress toward water and energy sustainability though enhanced water protection and conservation thus enhancing the viability of the domestic fossil fuel industry. The project contains 4 major independent Tasks. Task 1 - Work Plan: Water-Energy Sustainability: A Symposium on Resource Viability. Task 2 - Work Plan: A Regional Assessment of Water and Energy Sustainability. Task 3 - Work Plan: Risk Based Data Management System-Water Water and Energy Module. Task 4 - Work Plan: Identification and Assessment of States Regulatory Programs Regarding Geothermal Heating and Cooling Systems. Each task has a specific scope (details given).

  12. Re-water: More complicated than just toilet-to-tap, water reclamation helps sustain thirsty cities 

    E-Print Network [OSTI]

    Lee, Leslie

    2011-01-01

    stream_source_info Re-water_more complicated than just toilet-to-tap, water reclamation helps sustain thirsty cities.pdf.txt stream_content_type text/plain stream_size 14065 Content-Encoding ISO-8859-1 stream_name Re-water..._more complicated than just toilet-to-tap, water reclamation helps sustain thirsty cities.pdf.txt Content-Type text/plain; charset=ISO-8859-1 Fall 2011 tx H2O 25 Story by Leslie Lee Timeline of Droughts in Texas TWDB adopts Water for Texas 2007...

  13. Microheterogeneous Thoria-Urania Fuels for Pressurized Water Reactors

    SciTech Connect (OSTI)

    Shwageraus, Eugene; Zhao Xianfeng; Driscoll, Michael J.; Hejzlar, Pavel; Kazimi, Mujid S.; Herring, J. Stephen

    2004-07-15

    A thorium-based fuel cycle for light water reactors will reduce the plutonium generation rate and enhance the proliferation resistance of the spent fuel. However, priming the thorium cycle with {sup 235}U is necessary, and the {sup 235}U fraction in the uranium must be limited to below 20% to minimize proliferation concerns. Thus, a once-through thorium-uranium dioxide (ThO{sub 2}-UO{sub 2}) fuel cycle of no less than 25% uranium becomes necessary for normal pressurized water reactor (PWR) operating cycle lengths. Spatial separation of the uranium and thorium parts of the fuel can improve the achievable burnup of the thorium-uranium fuel designs through more effective breeding of {sup 233}U from the {sup 232}Th. Focus is on microheterogeneous fuel designs for PWRs, where the spatial separation of the uranium and thorium is on the order of a few millimetres to a few centimetres, including duplex pellet, axially microheterogeneous fuel, and a checkerboard of uranium and thorium pins. A special effort was made to understand the underlying reactor physics mechanisms responsible for enhancing the achievable burnup at spatial separation of the two fuels. The neutron spectral shift was identified as the primary reason for the enhancement of burnup capabilities. Mutual resonance shielding of uranium and thorium is also a factor; however, it is small in magnitude. It is shown that the microheterogeneous fuel can achieve higher burnups, by up to 15%, than the reference all-uranium fuel. However, denaturing of the {sup 233}U in the thorium portion of the fuel with small amounts of uranium significantly impairs this enhancement. The denaturing is also necessary to meet conventional PWR thermal limits by improving the power share of the thorium region at the beginning of fuel irradiation. Meeting thermal-hydraulic design requirements by some of the microheterogeneous fuels while still meeting or exceeding the burnup of the all-uranium case is shown to be potentially feasible. However, the large power imbalance between the uranium and thorium regions creates several design challenges, such as higher fission gas release and cladding temperature gradients. A reduction of plutonium generation by a factor of 3 in comparison with all-uranium PWR fuel using the same initial {sup 235}U content was estimated. In contrast to homogeneously mixed U-Th fuel, microheterogeneous fuel has a potential for economic performance comparable to the all-UO{sub 2} fuel provided that the microheterogeneous fuel incremental manufacturing costs are negligibly small.

  14. Sustainable Practices Policy Sections II, III.I. and V.I. Sustainable Water Systems

    E-Print Network [OSTI]

    California at Santa Cruz, University of

    , or lubricating equipment. Purified Water: Water that is free of impurities such as microorganisms, particulate consumption because it contains objectionable pollution, contamination minerals or infective agents, including water quality standards for human consumption. Reclaimed or Recycled Water: Wastewater treated

  15. An Afternoon with Professor Hermanowicz: Exploring Sustainability and Water Filtration

    E-Print Network [OSTI]

    2015-01-01

    focuses on the process of water quality treatment so thatthe quality of stored water. In this interview, Dr.health, on providing safe water, sanitation, and cleaning

  16. Consortium for Advanced Simulation of Light Water Reactors (CASL...

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

    reactor physical phenomena using coupled multiphysics models. VERA also includes the software development environment and computational infrastructure needed for these...

  17. Consortium for Advanced Simulation of Light Water Reactors (CASL...

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

    to achieve challenge problem solutions A strong VERA infrastructure supporting software development, testing, and releases. Requirements Drivers Modeling of reactors...

  18. CONSORTIUM FOR ADVANCED SIMULATION OF LIGHT WATER REACTORS (CASL...

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

    Problem Integration Virtual Reactor Integration (VRI) Bridging the gap between research and engineering. Chemistry Mesh Motion Quality Improvement Multi- resolution...

  19. pre-acts -6th annual international conference of Territorial Intelligence -caENTI October 2008 1 SUSTAINABLE WATER MANAGEMENT METHODS IN HUNGARY

    E-Print Network [OSTI]

    Boyer, Edmond

    a useful sustainable water management model based on the rainwater harvesting practices. Key words: Rainwater, sustainable water management, drinking water, rainwater harvesting, cistern, precipitation practical way to meet our everyday water needs is rainwater harvesting. Practically, this water is free

  20. Accident management for indian pressurized heavy water reactors

    SciTech Connect (OSTI)

    Hajela, S.; Grover, R.; Ghadge, S.G.; Bajaj, S.S. [Directorate of Safety, Nuclear Power Corporation of India Limited Nabhikiya Urja Bhawan, Anushakti Nagar, Mumbai-400 094 (India)

    2006-07-01

    Indian nuclear power program as of now is mainly based on Pressurized Heavy Water Reactors (PHWRs). Operating Procedures for normal power operation and Emergency Operating Procedures for operational transients and accidents within design basis exist for all Indian PHWRs. In addition, on-site and off-site emergency response procedures are also available for these NPPs. The guidelines needed for severe accidents mitigation are now formally being documented for Indian PHWRs. Also, in line with International trend of having symptom based emergency handling, the work is in advanced stage for preparation of symptom-based emergency operating procedures. Following a plant upset condition; a number of alarms distributed in different information systems appear in the control room to aid operator to identify the nature of the event. After identifying the event, appropriate intervention in the form of event based emergency operating procedure is put into use by the operating staff. However, if the initiating event cannot be unambiguously identified or after the initial event some other failures take place, then the selected event based emergency operating procedure will not be optimal. In such a case, reactor safety is ensured by monitoring safety functions (depicted by selected plant parameters grouped together) throughout the event handling so that the barriers to radioactivity release namely, fuel and fuel cladding, primary heat transport system integrity and containment remain intact. Simultaneous monitoring of all these safety functions is proposed through status trees and this concept will be implemented through a computer-based system. For beyond design basis accidents, event sequences are identified which may lead to severe core damage. As part of this project, severe accident mitigation guidelines are being finalized for the selected event sequences. The paper brings out the details of work being carried out for Indian PHWRs for symptom based event handling and severe accident management. (authors)

  1. Selection of a suitable reactor type for water desalination and power generation in Saudi Arabia

    SciTech Connect (OSTI)

    Hussein, F.M.

    1988-03-01

    Selection of a reactor type suitable for water desalination and power generation is a complex process that involves the evaluation of many criteria and requires the professional judgment of many experts in different fields. A reactor type that is suitable for one country might not be suitable for another. This is especially true in the case of Saudi Arabia because of its strategic location, the nature of its land and people, and its moderate technological situation. A detailed study using a computer code based on Saaty's mathematical pairwise comparison technique and developed in a previous study was carried out to find the most suitable reactor for water desalination and power generation in Saudi Arabia from among five potential types: boiling water reactors (BWRs), pressurized water reactors, CANDU heavy water reactors (HWRs), steam-generating heavy water reactors (SGHWRs), and high-temperature gas-cooled reactors. It was concluded that the CANDU HWR is the most suitable type for this purpose followed first by the BWR, then the SGHWR.

  2. Water Use and Sustainability in La Altagracia, Dominican Republic Caitlin Grady

    E-Print Network [OSTI]

    Beex, A. A. "Louis"

    Water Use and Sustainability in La Altagracia, Dominican Republic By Caitlin Grady Tamim Younos Research Experiences for Undergraduates (NSF REU) program Grant No.0649070, the Virginia Water Resources as necessarily representing the official policies, either expressed or implied, of the Virginia Water Resources

  3. Passive decay heat removal system for water-cooled nuclear reactors

    DOE Patents [OSTI]

    Forsberg, Charles W. (Oak Ridge, TN)

    1991-01-01

    A passive decay-heat removal system for a water-cooled nuclear reactor employs a closed heat transfer loop having heat-exchanging coils inside an open-topped, insulated box located inside the reactor vessel, below its normal water level, in communication with a condenser located outside of containment and exposed to the atmosphere. The heat transfer loop is located such that the evaporator is in a position where, when the water level drops in the reactor, it will become exposed to steam. Vapor produced in the evaporator passes upward to the condenser above the normal water level. In operation, condensation in the condenser removes heat from the system, and the condensed liquid is returned to the evaporator. The system is disposed such that during normal reactor operations where the water level is at its usual position, very little heat will be removed from the system, but during emergency, low water level conditions, substantial amounts of decay heat will be removed.

  4. Socio-hydrology: Use-inspired water sustainability science for the Anthropocene

    E-Print Network [OSTI]

    Sivapalan, M.

    Water is at the core of the most difficult sustainability challenges facing humans in the modern era, involving feedbacks across multiple scales, sectors, and agents. We suggest that a transformative new discipline is ...

  5. Multi-Application Small Light Water Reactor Final Report

    SciTech Connect (OSTI)

    Modro, S.M.; Fisher, J.E.; Weaver, K.D.; Reyes, J.N.; Groome, J.T.; Babka, P.; Carlson, T.M.

    2003-12-01

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle. Development of the baseline design concept has been sufficiently completed to determine that it complies with the safety requirements and criteria, and satisfies the major goals already noted. The more significant features of the baseline single-unit design concept include: (1) Thermal Power--150 MWt; (2) Net Electrical Output--35 MWe; (3) Steam Generator Type--Vertical, helical tubes; (4) Fuel UO{sub 2}, 8% enriched; (5) Refueling Intervals--5 years; (6) Life-Cycle--60 years. The economic performance was assessed by designing a power plant with an electric generation capacity in the range of current and advanced evolutionary systems. This approach allows for direct comparison of economic performance and forms a basis for further evaluation, economic and technical, of the proposed design and for the design evolution towards a more cost competitive concept. Applications such as cogeneration, water desalination or district heating were not addressed directly in the economic analyses since these depend more on local conditions, demand and economy and can not be easily generalized. Current economic performance experience and available cost data were used. The preliminary cost estimate, based on a concept that could be deployed in less than a decade, is: (1) Net Electrical Output--1050 MWe; (2) Net Station Efficiency--23%; (3) Number of Power Units--30; (4) Nominal Plant Capacity Factor--95%; (5) Total capital cost--$1241/kWe; and (6) Total busbar cost--3.4 cents/kWh. The project includes a testing program that has been conducted at Oregon State University (OSU). The test facility is a 1/3-height and 1/254.7 volume scaled design that will operate at full system pressure and temperature, and will be capable of operation at 600 kW. The design and construction of the facility have been completed. Testing is scheduled to begin in October 2002. The MASLWR conceptual design is simple, safe, and economical. It operates at NSSS parameters much lower than for a typical PWR plant, and has a much simplified power generation system. The individual reactor modules can be operated as on/off units, thereby limiting operational transients to startup and shutdown. In addition, a plant can be built in increments that match demand increases. The ''pull and replace'' concept offers automation of refueling and maintenance activities. Performing refueling in a single location improves proliferation resistance and eliminates the threat of diversion. Design certification based on testing is simplified because of the relatively low cost of a full-scale prototype facility. The overall conclusion is that while the efficiency of the power generation unit is much lower (23% versus 30%), the reduction in capital cost due to simplification of design more than makes up for the increased cost of nuclear fuel. The design concept complies with the safety requirements and criteria. It also satisfies the goals for modularity, standard plant design, certification before construction, c

  6. Multi-cycle boiling water reactor fuel cycle optimization

    SciTech Connect (OSTI)

    Ottinger, K.; Maldonado, G.I.

    2013-07-01

    In this work a new computer code, BWROPT (Boiling Water Reactor Optimization), is presented. BWROPT uses the Parallel Simulated Annealing (PSA) algorithm to solve the out-of-core optimization problem coupled with an in-core optimization that determines the optimum fuel loading pattern. However it uses a Haling power profile for the depletion instead of optimizing the operating strategy. The result of this optimization is the optimum new fuel inventory and the core loading pattern for the first cycle considered in the optimization. Several changes were made to the optimization algorithm with respect to other nuclear fuel cycle optimization codes that use PSA. Instead of using constant sampling probabilities for the solution perturbation types throughout the optimization as is usually done in PSA optimizations the sampling probabilities are varied to get a better solution and/or decrease runtime. The new fuel types available for use can be sorted into an array based on any number of parameters so that each parameter can be incremented or decremented, which allows for more precise fuel type selection compared to random sampling. Also, the results are sorted by the new fuel inventory of the first cycle for ease of comparing alternative solutions. (authors)

  7. Experimental Studies of NGNP Reactor Cavity Cooling System With Water

    SciTech Connect (OSTI)

    Corradini, Michael; Anderson, Mark; Hassan, Yassin; Tokuhiro, Akira

    2013-01-16

    This project will investigate the flow behavior that can occur in the reactor cavity cooling system (RCCS) with water coolant under the passive cooling-mode of operation. The team will conduct separate-effects tests and develop associated scaling analyses, and provide system-level phenomenological and computational models that describe key flow phenomena during RCCS operation, from forced to natural circulation, single-phase flow and two-phase flow and flashing. The project consists of the following tasks: Task 1. Conduct separate-effects, single-phase flow experiments and develop scaling analyses for comparison to system-level computational modeling for the RCCS standpipe design. A transition from forced to natural convection cooling occurs in the standpipe under accident conditions. These tests will measure global flow behavior and local flow velocities, as well as develop instrumentation for use in larger scale tests, thereby providing proper flow distribution among standpipes for decay heat removal. Task 2. Conduct separate-effects experiments for the RCCS standpipe design as two-phase flashing occurs and flow develops. As natural circulation cooling continues without an ultimate heat sink, water within the system will heat to temperatures approaching saturation , at which point two-phase flashing and flow will begin. The focus is to develop a phenomenological model from these tests that will describe the flashing and flow stability phenomena. In addition, one could determine the efficiency of phase separation in the RCCS storage tank as the two-phase flashing phenomena ensues and the storage tank vents the steam produced. Task 3. Develop a system-level computational model that will describe the overall RCCS behavior as it transitions from forced flow to natural circulation and eventual two-phase flow in the passive cooling-mode of operation. This modeling can then be used to test the phenomenological models developed as a function of scale.

  8. Consortium for Advanced Simulation of Light Water Reactors (CASL...

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

    products; Develop and evolve within VERA physics-based materials models of the fuel system, reactor vessel and internals, improved constitutive relations of coolant and...

  9. Consortium for Advanced Simulation of Light Water Reactors (CASL...

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

    LWRs; Develop and effectively apply modern virtual reactor technology; Engage the nuclear energy community through modeling and simulation; and Deploy new partnership and...

  10. Consortium for Advanced Simulation of Light Water Reactors (CASL...

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

    Radiation Transport Methods (RTM) Delivers next-generation radiation transport tools to the virtual Reactor RTM Vision Statement Objectives and Strategies Next generation,...

  11. CONSORTIUM FOR ADVANCED SIMULATION OF LIGHT WATER REACTORS (CASL...

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

    strategic plan to ensure long term success, and he presented one of several roadmaps, Reactor Applications. One IC member commented that the roadmap should be changes so that...

  12. Light Water Reactors A DOE Energy Innovation Hub for Modeling...

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

    and Simulation of Nuclear Reactors CASL is focused on three issues for nuclear energy: reducing cost, reducing the amount of used nuclear fuel, and safety. CASL core...

  13. Consortium for Advanced Simulation of Light Water Reactors (CASL...

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

    Thermal Hydraulics Methods (THM) Delivers next-generation thermal-hydraulic simulation tools to Virtual Environment for Reactor Applications (VERA) Thermal Hydraulics Methods...

  14. Consortium for Advanced Simulation of Light Water Reactors (CASL...

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

    user forum User support and VERA distribution systems capable of meeting user demand A sustainable model for continued use and development of CASL's technology after the active...

  15. Analysis of the Simplified Boiling Water Reactor using the code Ramona-4B 

    E-Print Network [OSTI]

    Cuevas Vivas, Gabriel Francisco

    1995-01-01

    The analysis of the Simplified Boiling Water Reactor (SBVVR) is carried out through the use of the reactor analysis code RAMONA-4B in a scenario of an operational transient, a turbine trip with failure of all the bypass valves. This study is divided...

  16. Secondary Startup Neutron Sources as a Source of Tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS)

    SciTech Connect (OSTI)

    Shaver, Mark W.; Lanning, Donald D.

    2010-02-01

    The hypothesis of this paper is that the Zircaloy clad fuel source is minimal and that secondary startup neutron sources are the significant contributors of the tritium in the RCS that was previously assigned to release from fuel. Currently there are large uncertainties in the attribution of tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS). The measured amount of tritium in the coolant cannot be separated out empirically into its individual sources. Therefore, to quantify individual contributors, all sources of tritium in the RCS of a PWR must be understood theoretically and verified by the sum of the individual components equaling the measured values.

  17. Assessment of light water reactor power plant cost and ultra-acceleration depreciation financing

    E-Print Network [OSTI]

    El-Magboub, Sadek Abdulhafid.

    Although in many regions of the U.S. the least expensive electricity is generated from light-water reactor (LWR) plants, the fixed (capital plus operation and maintenance) cost has increased to the level where the cost ...

  18. Analysis of strategies for improving uranium utilization in pressurized water reactors

    E-Print Network [OSTI]

    Sefcik, Joseph A.

    1981-01-01

    Systematic procedures have been devised and applied to evaluate core design and fuel management strategies for improving uranium utilization in Pressurized Water Reactors operated on a once-through fuel cycle. A principal ...

  19. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    DOE Patents [OSTI]

    McDermott, D.J.; Schrader, K.J.; Schulz, T.L.

    1994-05-03

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  20. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    DOE Patents [OSTI]

    McDermott, Daniel J. (Export, PA); Schrader, Kenneth J. (Penn Hills, PA); Schulz, Terry L. (Murrysville Boro, PA)

    1994-01-01

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  1. EIS-0288: Production of Tritium in a Commercial Light Water Reactor

    Office of Energy Efficiency and Renewable Energy (EERE)

    This Environmental Impact Statement for the Production of Tritium in a Commercial Light Water Reactor (CLWR EIS) evaluates the environmental impacts associated with producing tritium at one or more...

  2. The selective use of thorium and heterogeneity in uranium-efficient pressurized water reactors

    E-Print Network [OSTI]

    Kamal, Altamash

    1982-01-01

    Systematic procedures have been developed and applied to assess the uranium utilization potential of a broad range of options involving the selective use of thorium in Pressurized Water Reactors (PWRs) operating on the ...

  3. An inverted pressurized water reactor design with twisted-tape swirl promoters

    E-Print Network [OSTI]

    Nguyen, Nghia T. (Nghia Tat)

    2014-01-01

    An Inverted Fuel Pressurized Water Reactor (IPWR) concept was previously investigated and developed by Paolo Ferroni at MIT with the effort to improve the power density and capacity of current PWRs by modifying the core ...

  4. Design strategies for optimizing high burnup fuel in pressurized water reactors

    E-Print Network [OSTI]

    Xu, Zhiwen, 1975-

    2003-01-01

    This work is focused on the strategy for utilizing high-burnup fuel in pressurized water reactors (PWR) with special emphasis on the full array of neutronic considerations. The historical increase in batch-averaged discharge ...

  5. CHEMICAL ASPECTS OF PELLET-CLADDING INTERACTION IN LIGHT WATER REACTOR FUEL ELEMENTS

    E-Print Network [OSTI]

    Olander, D.R.

    2010-01-01

    in LWR Fuel Pods Relevant to PCI Failures", paper presented1981). D. R. Olander, "Is PCI an Iodine SCC Phenomenon?" ,pellet-cladding interaction(PCI) in light water reactor fuel

  6. Feasibility of breeding in hard spectrum boiling water reactors with oxide and nitride fuels

    E-Print Network [OSTI]

    Feng, Bo, Ph. D. Massachusetts Institute of Technology

    2011-01-01

    This study assesses the neutronic, thermal-hydraulic, and fuel performance aspects of using nitride fuel in place of oxides in Pu-based high conversion light water reactor designs. Using the higher density nitride fuel ...

  7. Conceptual design of an annular-fueled superheat boiling water reactor

    E-Print Network [OSTI]

    Ko, Yu-Chih, Ph. D. Massachusetts Institute of Technology

    2011-01-01

    The conceptual design of an annular-fueled superheat boiling water reactor (ASBWR) is outlined. The proposed design, ASBWR, combines the boiler and superheater regions into one fuel assembly. This ensures good neutron ...

  8. Stability analysis of the boiling water reactor : methods and advanced designs

    E-Print Network [OSTI]

    Hu, Rui, Ph. D. Massachusetts Institute of Technology

    2010-01-01

    Density Wave Oscillations (DWOs) are known to be possible when a coolant undergoes considerable density reduction while passing through a heated channel. In the development of boiling water reactors (BWRs), there has been ...

  9. SWS 5246 Water Resource Sustainability 3 credits/Spring Semester

    E-Print Network [OSTI]

    Ma, Lena

    ) 6 Water transfers (including flow diversion and transboundary management) 7 - 10 Hydraulic infrastructure (including canals and dams) 11-12 Water institutions (including large-scale management) 13-15 #12

  10. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production

    SciTech Connect (OSTI)

    Philip MacDonald; Jacopo Buongiorno; James Sterbentz; Cliff Davis; Robert Witt; Gary Was; J. McKinley; S. Teysseyre; Luca Oriani; Vefa Kucukboyaci; Lawrence Conway; N. Jonsson: Bin Liu

    2005-02-13

    The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency. It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor.

  11. Innovative fuel designs for high power density pressurized water reactor

    E-Print Network [OSTI]

    Feng, Dandong, Ph. D. Massachusetts Institute of Technology

    2006-01-01

    One of the ways to lower the cost of nuclear energy is to increase the power density of the reactor core. Features of fuel design that enhance the potential for high power density are derived based on characteristics of ...

  12. An integrated approach for the verification of fresh mixed oxide fuel (MOX) assemblies at light water reactor MOX recycle reactors

    SciTech Connect (OSTI)

    Menlove, Howard O; Lee, Sang - Yoon

    2009-01-01

    This paper presents an integrated approach for the verification of mixed oxide (MOX) fuel assemblies prior to their being loaded into the reactor. There is a coupling of the verification approach that starts at the fuel fabrication plant and stops with the transfer of the assemblies into the thermal reactor. The key measurement points are at the output of the fuel fabrication plant, the receipt at the reactor site, and the storage in the water pool as fresh fuel. The IAEA currently has the capability to measure the MOX fuel assemblies at the output of the fuel fabrication plants using a passive neutron coincidence counting systems of the passive neutron collar (PNCL) type. Also. at the MOX reactor pool, the underwater coincidence counter (UWCC) has been developed to measure the MOX assemblies in the water. The UWCC measurement requires that the fuel assembly be lifted about two meters up in the storage rack to avoid interference from the fuel that is stored in the rack. This paper presents a new method to verify the MOX fuel assemblies that are in the storage rack without the necessity of moving the fuel. The detector system is called the Underwater MOX Verification System (UMVS). The integration and relationship of the three measurements systems is described.

  13. Integration of Nutrient and Water Recycling for Sustainable Algal...

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

    water-recovery methods. Characterize the development, structure, and stability of microbial communities in algal systems. Perform economic and life cycle assessments...

  14. Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

    SciTech Connect (OSTI)

    1993-09-15

    This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

  15. Comparative assessment of nuclear fuel cycles. Light-water reactor once-through, classical fast breeder reactor, and symbiotic fast breeder reactor cycles

    SciTech Connect (OSTI)

    Hardie, R.W.; Barrett, R.J.; Freiwald, J.G.

    1980-06-01

    The object of the Alternative Nuclear Fuel Cycle Study is to perform comparative assessments of nuclear power systems. There are two important features of this study. First, this evaluation attempts to encompass the complete, integrated fuel cycle from mining of uranium ore to disposal of waste rather than isolated components. Second, it compares several aspects of each cycle - energy use, economics, technological status, proliferation, public safety, and commercial potential - instead of concentrating on one or two assessment areas. This report presents assessment results for three fuel cycles. These are the light-water reactor once-through cycle, the fast breeder reactor on the classical plutonium cycle, and the fast breeder reactor on a symbiotic cycle using plutonium and /sup 233/U as fissile fuels. The report also contains a description of the methodology used in this assessment. Subsequent reports will present results for additional fuel cycles.

  16. The National Hydropower Asset Assessment Program (NHAAP) is an integrated energy, water, and ecosystem research effort for sustainable hydroelectricity generation and water management. The NHAAP conducts research on new

    E-Print Network [OSTI]

    , and ecosystem research effort for sustainable hydroelectricity generation and water management. The NHAAP

  17. Talking sustainability:Federal initiatives target major water resources concerns 

    E-Print Network [OSTI]

    Lee, Leslie

    2011-01-01

    . The following Texas Water Resources Institute (TWRI) projects have impacted major water resource problems in Texas and beyond. Ogallala Aquifer Project The Ogallala Aquifer, stretching from South Dakota to Texas, covers 174,000 square miles, including... 36,080 square miles in the Texas High Plains. In western Kansas and the Texas High Plains, the aquifer is declining at an unacceptable rate. Aquifer depletion rates of 1 to 3 feet per year are common in that region, and recharges are very...

  18. Talking sustainability: Federal intiatives target major water resources concerns 

    E-Print Network [OSTI]

    Lee, Leslie

    2010-01-01

    . The following Texas Water Resources Institute (TWRI) projects have impacted major water resource problems in Texas and beyond. Ogallala Aquifer Project The Ogallala Aquifer, stretching from South Dakota to Texas, covers 174,000 square miles, including... 36,080 square miles in the Texas High Plains. In western Kansas and the Texas High Plains, the aquifer is declining at an unacceptable rate. Aquifer depletion rates of 1 to 3 feet per year are common in that region, and recharges are very...

  19. Sustainability

    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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power AdministrationRobust,Field-effectWorking With Livermore NationalSurprisingSustainability Sign In About |

  20. Nuclear reactor with makeup water assist from residual heat removal system

    DOE Patents [OSTI]

    Corletti, Michael M. (New Kensington, PA); Schulz, Terry L. (Murrysville, PA)

    1993-01-01

    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path.

  1. Nuclear reactor with makeup water assist from residual heat removal system

    DOE Patents [OSTI]

    Corletti, M.M.; Schulz, T.L.

    1993-12-07

    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures.

  2. FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

    SciTech Connect (OSTI)

    Greene, S.R.; Spellman, D.J.; Bevard, B.B. [and others

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative.

  3. Comparison of actinide production in traveling wave and pressurized water reactors

    SciTech Connect (OSTI)

    Osborne, A.G.; Smith, T.A.; Deinert, M.R.

    2013-07-01

    The geopolitical problems associated with civilian nuclear energy production arise in part from the accumulation of transuranics in spent nuclear fuel. A traveling wave reactor is a type of breed-burn reactor that could, if feasible, reduce the overall production of transuranics. In one possible configuration, a cylinder of natural or depleted uranium would be subjected to a fast neutron flux at one end. The neutrons would transmute the uranium, producing plutonium and higher actinides. Under the right conditions, the reactor could become critical, at which point a self-stabilizing fission wave would form and propagate down the length of the reactor cylinder. The neutrons from the fission wave would burn the fissile nuclides and transmute uranium ahead of the wave to produce additional fuel. Fission waves in uranium are driven largely by the production and fission of {sup 239}Pu. Simulations have shown that the fuel burnup can reach values greater than 400 MWd/kgIHM, before fission products poison the reaction. In this work we compare the production of plutonium and minor actinides produced in a fission wave to that of a UOX fueled light water reactor, both on an energy normalized basis. The nuclide concentrations in the spent traveling wave reactor fuel are computed using a one-group diffusion model and are verified using Monte Carlo simulations. In the case of the pressurized water reactor, a multi-group collision probability model is used to generate the nuclide quantities. We find that the traveling wave reactor produces about 0.187 g/MWd/kgIHM of transuranics compared to 0.413 g/MWd/kgIHM for a pressurized water reactor running fuel enriched to 4.95 % and burned to 50 MWd/kgIHM. (authors)

  4. Jess S. Jackson Sustainable Winery Building WATER STORAGE

    E-Print Network [OSTI]

    Ullrich, Paul

    at adjacent winery to irrigate the landscape SOLAR ARRAY produces enough energy to offset the building Certification, the facility is designed for net-zero energy, carbon, and water thresholds, and houses innovative panels to offset future energy usage of process equipment installed in the building by the owner

  5. Energy/Water Sustainability and the Electric Power

    E-Print Network [OSTI]

    Keller, Arturo A.

    are using wet cooling tower) Water Use by Plant Type 0 100 200 300 400 500 600 700 800 900 Nuclear Coal Oil Power Research Institute, Inc. All rights reserved. Dry Cooling Matimba 6x665MW Coal Courtesy of Eskom

  6. Tritium recovery from tritiated water with a two-stage palladium membrane reactor

    SciTech Connect (OSTI)

    Birdsell, S.A.; Willms, R.S.

    1997-04-01

    A process to recover tritium from tritiated water has been successfully demonstrated at TSTA. The 2-stage palladium membrane reactor (PMR) is capable of recovering tritium from water without generating additional waste. This device can be used to recover tritium from the substantial amount of tritiated water that is expected to be generated in the International Thermonuclear Experimental Reactor both from torus exhaust and auxiliary operations. A large quantity of tritiated waste water exists world wide because the predominant method of cleaning up tritiated streams is to oxidize tritium to tritiated water. The latter can be collected with high efficiency for subsequent disposal. The PMR is a combined catalytic reactor/permeator. Cold (non-tritium) water processing experiments were run in preparation for the tritiated water processing tests. Tritium was recovered from a container of molecular sieve loaded with 2,050 g (2,550 std. L) of water and 4.5 g of tritium. During this experiment, 27% (694 std. L) of the water was processed resulting in recovery of 1.2 g of tritium. The maximum water processing rate for the PMR system used was determined to be 0.5 slpm. This correlates well with the maximum processing rate determined from the smaller PMR system on the cold test bench and has resulted in valuable scale-up and design information.

  7. Comparative study of plutonium burning in heavy and light water reactors.

    SciTech Connect (OSTI)

    Taiwo, T. A.; Kim, T. K.; Szakaly, F. J.; Hill, R. N.; Yang, W. S.; Dyck, G. R.; Hyland, B.; Edwards, G. W. R.; Nuclear Engineering Division; Atomic Energy Canada Ltd.

    2008-01-01

    There is interest in the U.S. and world-wide in reducing the burden on geological nuclear fuel disposal sites. In some disposal scenarios, the decay heat loading of the surrounding rock limits the commercial spent fuel capacity of the sites. In the long term (100 to 1,500 years), this decay heat is generated primarily by actinides, particularly {sup 241}Am and {sup 241}Pu. One possible approach to reducing this decay-heat burden would be to reprocess commercial spent nuclear fuel and use intermediate-tier thermal reactors to 'burn' these actinides and other transuranics (plutonium and higher actinides). The viability of this approach is dependent on the detailed changes in chemical and isotopic compositions of actinide-bearing fuels after irradiation in thermal reactor spectra. The intermediate-tier thermal burners could bridge the commercial water-cooled reactors and fast reactors required for ultimate consumption of the transuranics generated in the commercial reactors. This would reduce the number of such fast reactors required to complete the mission of burning transuranics. If thermal systems are to be used for the transmutation mission, it is likely that they would be similar to or are advanced versions of the systems currently used for power generation. In both the U.S. and Canada, light- and heavy-water-cooled thermal reactors are used for power generation in the commercial nuclear sector. About 103 pressurized- and boiling- light water reactors (PWRs and BRWs) are deployed in the U.S. nuclear industry while about 18 CANDU (heavy-water-cooled) reactors are used in the Canadian industry. There are substantial differences between light and heavy water-cooled reactors that might affect transmutation potential. These arise from differences in neutron balance of the reactors, in neutron energy spectra, in operational approaches (e.g., continuous refueling enhancing fuel burnup), and so on. A systematic study has been conducted to compare the transmutation potentials of CANDU and PWR systems using (U,Pu)O{sub 2} mixed oxide fuels. First, we examine and compare the isotopic evolution of plutonium-containing fuel under irradiation in these reactor types to understand the physics processes involved. The core-physics parameters to be compared for these systems are generated using two-dimensional lattice physics models for a single fuel assembly that is representative of the whole-core (e.g., using the linear reactivity model). Results from a parametric study of the discharge burnup as a function of the Pu fraction in the initial heavy metal are presented for each system. The Pu consumption level, minor actinides buildup level, and the masses destroyed per unit energy generation are summarized and compared. In addition, assessment results for a simple plutonium recycling concept in realistic CANDU and PWR cores are presented. In these cases, plutonium from commercial spent PWR fuel will be separated and burned in realistic intermediate thermal burner reactors using (U,Pu)O{sub 2} mixed oxide fuel. The spent fuel from this thermal burner will be separated and the resulting Pu will be burned in a second pass through the thermal burner reactor. The resulting transuranics are assumed to then be burned in a fast burner reactor. The impact of using the spent fuels of these systems on the core performance of the fast burner reactor and the required numbers of the various reactor types will be discussed.

  8. Evolutionary/advanced light water reactor data report

    SciTech Connect (OSTI)

    1996-02-09

    The US DOE Office of Fissile Material Disposition is examining options for placing fissile materials that were produced for fabrication of weapons, and now are deemed to be surplus, into a condition that is substantially irreversible and makes its use in weapons inherently more difficult. The principal fissile materials subject to this disposition activity are plutonium and uranium containing substantial fractions of plutonium-239 uranium-235. The data in this report, prepared as technical input to the fissile material disposition Programmatic Environmental Impact Statement (PEIS) deal only with the disposition of plutonium that contains well over 80% plutonium-239. In fact, the data were developed on the basis of weapon-grade plutonium which contains, typically, 93.6% plutonium-239 and 5.9% plutonium-240 as the principal isotopes. One of the options for disposition of weapon-grade plutonium being considered is the power reactor alternative. Plutonium would be fabricated into mixed oxide (MOX) fuel and fissioned (``burned``) in a reactor to produce electric power. The MOX fuel will contain dioxides of uranium and plutonium with less than 7% weapon-grade plutonium and uranium that has about 0.2% uranium-235. The disposition mission could, for example, be carried out in existing power reactors, of which there are over 100 in the United States. Alternatively, new LWRs could be constructed especially for disposition of plutonium. These would be of the latest US design(s) incorporating numerous design simplifications and safety enhancements. These ``evolutionary`` or ``advanced`` designs would offer not only technological advances, but also flexibility in siting and the option of either government or private (e.g., utility) ownership. The new reactor designs can accommodate somewhat higher plutonium throughputs. This data report deals solely with the ``evolutionary`` LWR alternative.

  9. 309NUCLEAR ENGINEERING AND TECHNOLOGY, VOL.37 NO.4, AUGUST 2005 A NEW BOOK: "LIGHT-WATER REACTOR MATERIALS"

    E-Print Network [OSTI]

    Motta, Arthur T.

    309NUCLEAR ENGINEERING AND TECHNOLOGY, VOL.37 NO.4, AUGUST 2005 A NEW BOOK: "LIGHT-WATER REACTOR review; it is a book preview. Thirty years ago, "Fundamental Aspects of Nuclear Reactor Fuel Elements of nuclear fuels among other topics pertinent to the materials in the ensemble of the nuclear reactor

  10. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production

    SciTech Connect (OSTI)

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

    2002-01-01

    The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

  11. Design and Testing of Vacuum Breaker Check Valve for Simplified Boiling Water Reactor

    SciTech Connect (OSTI)

    Ishii, M.; Xu, Y.; Revankar, S.T.

    2002-07-01

    A new design of the vacuum breaker check valve was developed to replace the mechanical valve in a simplified boiling water reactor. Scaling and design calculations were performed to obtain the geometry of new passive hydraulic vacuum breaker check valve. In order to check the valve performance, a RELAP5 model of the simplified boiling water reactor system with the new valve was developed. The valve was implemented in an integral facility, PUMA and was tested for large break loss of coolant accident. (authors)

  12. Sustaining Industrial Energy Efficiency in Process Cooling in a Potentially Water-Short Future 

    E-Print Network [OSTI]

    Ferland, K.

    2014-01-01

    stream_source_info ESL-IE-14-05-18.pdf.txt stream_content_type text/plain stream_size 14223 Content-Encoding UTF-8 stream_name ESL-IE-14-05-18.pdf.txt Content-Type text/plain; charset=UTF-8 Sustaining Industrial Energy... and Management • Water Reuse and Use of Unconventional Sources ESL-IE-14-05-18 Proceedings of the Thrity-Sixth Industrial Energy Technology Conference New Orleans, LA. May 20-23, 2014 Topic 1: Existing Technology and Operations Strategy: Sustain process...

  13. Component failures at pressurized water reactors. Final report

    SciTech Connect (OSTI)

    Reisinger, M.F.

    1980-10-01

    Objectives of this study were to identify those systems having major impact on safety and availability (i.e. to identify those systems and components whose failures have historically caused the greatest number of challenges to the reactor protective systems and which have resulted in greatest loss of electric generation time). These problems were identified for engineering solutions and recommendations made for areas and programs where research and development should be concentrated. The program was conducted in three major phases: Data Analysis, Engineering Evaluation, Cost Benefit Analysis.

  14. A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 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.

    2010-01-01

    Charges Relating to Nuclear Reactor Safety," 1976, availablestudies of light-water nuclear reactor safety, emphasizingstudies of overall nuclear reactor safety have been

  15. A Qualitative Assessment of Thorium-Based Fuels in Supercritical Pressure Water Cooled Reactors

    SciTech Connect (OSTI)

    Weaver, Kevan Dean; Mac Donald, Philip Elsworth

    2002-10-01

    The requirements for the next generation of reactors include better economics and safety, waste minimization (particularly of the long-lived isotopes), and better proliferation resistance (both intrinsic and extrinsic). A supercritical pressure water cooled reactor has been chosen as one of the lead contenders as a Generation IV reactor due to the high thermal efficiency and compact/simplified plant design. In addition, interest in the use of thorium-based fuels for Generation IV reactors has increased based on the abundance of thorium, and the minimization of transuranics in a neutron flux; as plutonium (and thus the minor actinides) is not a by-product in the thorium chain. In order to better understand the possibility of the combination of these concepts to meet the Generation IV goals, the qualitative burnup potential and discharge isotopics of thorium and uranium fuel were studied using pin cell analyses in a supercritical pressure water cooled reactor environment. Each of these fertile materials were used in both nitride and metallic form, with light water reactor grade plutonium and minor actinides added. While the uranium-based fuels achieved burnups that were 1.3 to 2.7 times greater than their thorium-based counterparts, the thorium-based fuels destroyed 2 to 7 times more of the plutonium and minor actinides. The fission-to-capture ratio is much higher in this reactor as compared to PWR’s and BWR’s due to the harder neutron spectrum, thus allowing more efficient destruction of the transuranic elements. However, while the uranium-based fuels do achieve a net depletion of plutonium and minor actinides, the breeding of these isotopes limits this depletion; especially as compared to the thorium-based fuels.

  16. Nanostructure of Metallic Particles in Light Water Reactor Used Nuclear Fuel

    SciTech Connect (OSTI)

    Buck, Edgar C.; Mausolf, Edward J.; Mcnamara, Bruce K.; Soderquist, Chuck Z.; Schwantes, Jon M.

    2015-03-11

    The extraordinary nano-structure of metallic particles in light water reactor fuels points to possible high reactivity through increased surface area and a high concentration of high energy defect sites. We have analyzed the metallic epsilon particles from a high burn-up fuel from a boiling water reactor using transmission electron microscopy and have observed a much finer nanostructure in these particles than has been reported previously. The individual round particles that varying in size between ~20 and ~50 nm appear to consist of individual crystallites on the order of 2-3 nm in diameter. It is likely that in-reactor irradiation induce displacement cascades results in the formation of the nano-structure. The composition of these metallic phases is variable yet the structure of the material is consistent with the hexagonal close packed structure of epsilon-ruthenium. These findings suggest that unusual catalytic behavior of these materials might be expected, particularly under accident conditions.

  17. Application of the Isotope Ratio Method to a Boiling Water Reactor

    SciTech Connect (OSTI)

    Frank, Douglas P.; Gerlach, David C.; Gesh, Christopher J.; Hurley, David E.; Meriwether, George H.; Mitchell, Mark R.; Reid, Bruce D.

    2010-08-11

    The isotope ratio method is a technique for estimating the energy or plutonium production in a fission reactor by measuring isotope ratios in non-fuel reactor components. The isotope ratios in these components can then be directly related to the cumulative energy production with standard reactor modeling methods. All reactor materials contain trace elemental impurities at parts per million levels, and the isotopes of these elements are transmuted by neutron irradiation in a predictable manner. While measuring the change in a particular isotope’s concentration is possible, it is difficult to correlate to energy production because the initial concentration of that element may not be accurately known. However, if the ratio of two isotopes of the same element can be measured, the energy production can then be determined without knowing the absolute concentration of that impurity since the initial natural ratio is known. This is the fundamental principle underlying the isotope ratio method. Extremely sensitive mass-spectrometric methods are currently available that allow accurate measurements of the impurity isotope ratios in samples. Additionally, “indicator” elements with stable activation products have been identified so that their post-irradiation isotope ratios remain constant. This method has been successfully demonstrated on graphite-moderated reactors. Graphite reactors are particularly well-suited to such analyses since the graphite moderator is resident in the fueled region of the core for the entire period of operation. Applying this method to other reactor types is more difficult since the resident portions of the reactor available for sampling are either outside the fueled region of the core or structural components of individual fuel assemblies. The goal of this research is to show that the isotope ratio method can produce meaningful results for light water-moderated power reactors. In this work, we use the isotope ratio method to estimate the energy production in a boiling water reactor fuel bundle based on measurements taken from the corresponding fuel assembly channel. Our preliminary results are in good agreement with the actual operating history of the reactor during the cycle that the fuel bundle was resident in the core.

  18. iUTAH Water Sustainability Graduate Research Fellows: Call for Proposals The NSF EPSCoR innovative Urban Transitions and Arid-region Hydro-sustainability (iUTAH)

    E-Print Network [OSTI]

    Tipple, Brett

    R innovative Urban Transitions and Arid-region Hydro-sustainability (iUTAH) program announces a 2012-2013 call1 iUTAH Water Sustainability Graduate Research Fellows: Call for Proposals The NSF EPSCo broadly pertaining to water sustainability in the Wasatch Front are eligible to apply. This program

  19. Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

  20. A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 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.

    2010-01-01

    this man- (PWRs) and boiling-water reactors to compare thesepressur- ized and boiling water reactors is presented. ItAmerican Physical reactor cooling - boiling-water ECC(S) -

  1. In-Reactor Oxidation of Zircaloy-4 Under Low Water Vapor Pressures

    SciTech Connect (OSTI)

    Luscher, Walter G.; Senor, David J.; Clayton, Kevin; Longhurst, Glen

    2015-01-01

    Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330° and 370°C). Data from these tests will be used to support fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr-4 over the specified range of test conditions. Comparisons between in- and ex- reactor test results were performed to evaluate the influence of irradiation.

  2. In-reactor oxidation of zircaloy-4 under low water vapor pressures

    SciTech Connect (OSTI)

    Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

    2015-01-01

    Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 ºC). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr- 4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

  3. Interactions of woody biofuel feedstock production systems with water resources: Considerations for sustainability.

    SciTech Connect (OSTI)

    Trettin, Carl,C.; Amatya, Devendra; Coleman, Mark.

    2008-07-01

    Abstract. Water resources are important for the production of woody biofuel feedstocks. It is necessary to ensure that production systems do not adversely affect the quantity or quality of surface and ground water. The effects of woody biomass plantations on water resources are largely dependent on the prior land use and the management regime. Experience from both irrigated and non-irrigated systems has demonstrated that woody biofuel production systems do not impair water quality. Water quality actually improves from conversion of idle or degraded agricultural lands to woody biomass plantations. Site water balance may be altered by cultivation of woody biomass plantations relative to agricultural use, due to increases in evapostranspiration (ET) and storage. Incorporation of woody biomass production plantations within the landscape provides an opportunity to improve the quality of runoff water and soil conservation. Given the centrality of water resources to the sustainability of ecosystem services and other values derived, the experience with woody biofuels feedstock production systems is positive. Keywords. Short rotation woody crop, forest hydrology, water quality, hardwood plantation.

  4. Environmentally assisted cracking in light water reactors annual report January - December 2005.

    SciTech Connect (OSTI)

    Alexandreanu, B.; Chen, Y.; Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.

    2007-08-31

    This report summarizes work performed from January to December 2005 by Argonne National Laboratory on fatigue and environmentally assisted cracking in light water reactors (LWRs). Existing statistical models for estimating the fatigue life of carbon and low-alloy steels and austenitic stainless steels (SSs) as a function of material, loading, and environmental conditions were updated. Also, the ASME Code fatigue adjustment factors of 2 on stress and 20 on life were critically reviewed to assess the possible conservatism in the current choice of the margins. An approach, based on an environmental fatigue correction factor, for incorporating the effects of LWR environments into ASME Section III fatigue evaluations is discussed. The susceptibility of austenitic stainless steels and their welds to irradiation-assisted stress corrosion cracking (IASCC) is being evaluated as a function of the fluence level, water chemistry, material chemistry, and fabrication history. For this task, crack growth rate (CGR) tests and slow strain rate tensile (SSRT) tests are being conducted on various austenitic SSs irradiated in the Halden boiling water reactor. The SSRT tests are currently focused on investigating the effects of the grain boundary engineering process on the IASCC of the austenitic SSs. The CGR tests were conducted on Type 316 SSs irradiated to 0.45-3.0 dpa, and on sensitized Type 304 SS and SS weld heat-affected-zone material irradiated to 2.16 dpa. The CGR tests on materials irradiated to 2.16 dpa were followed by a fracture toughness test in a water environment. The effects of material composition, irradiation, and water chemistry on growth rates are discussed. The susceptibility of austenitic SS core internals to IASCC and void swelling is also being evaluated for pressurized water reactors. Both SSRT tests and microstructural examinations are being conducted on specimens irradiated in the BOR-60 reactor in Russia to doses up to 20 dpa. Crack growth rate data, obtained in the pressurized water reactor environment, are presented on Ni-alloy welds prepared in the laboratory or obtained from the nozzle-to-pipe weld of the V. C. Summer reactor. The experimental CGRs under cyclic and constant load are compared with the existing CGR data for Ni-alloy welds to determine the relative susceptibility of these materials to environmentally enhanced cracking under a variety of loading conditions.

  5. Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2003-01-01

    1.1 This guide covers the general procedures to be considered for conducting an in-service thermal anneal of a light-water moderated nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. The improvement in mechanical properties generally is assessed using Charpy V-notch impact test results, or alternatively, fracture toughness test results or inferred toughness property changes from tensile, hardness, indentation, or other miniature specimen testing (1). 1.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods. Certain inherent limiting factors must be considered in developing an annealing procedure. These factors include system-design limitations; physical constrain...

  6. Passive and inherent safety technologies for light-water nuclear reactors

    SciTech Connect (OSTI)

    Forsberg, C.W.

    1990-07-01

    Passive/inherent safety implies a technical revolution in our approach to nuclear power safety. This direction is discussed herein for light-water reactors (LWRs) -- the predominant type of power reactor used in the world today. At Oak Ridge National Laboratory (ORNL) the approach to the development of passive/inherent safety for LWRs consists of four steps: identify and quantify safety requirements and goals; identify and quantify the technical functional requirements needed for safety; identify, invent, develop, and quantify technical options that meet both of the above requirements; and integrate safety systems into designs of economic and reliable nuclear power plants. Significant progress has been achieved in the first three steps of this program. The last step involves primarily the reactor vendors. These activities, as well as related activities worldwide, are described here. 27 refs., 7 tabs.

  7. Plutonium Recycling in Light Water Reactors at Framatome ANP: Status and Trends

    SciTech Connect (OSTI)

    Porsch, Dieter [Framatome ANP GmbH (France); Stach, Walter [Framatome ANP GmbH (France); Charmensat, Pascal [Framatome ANP S.A.S. (France); Pasquet, Michel [Framatome ANP S.A.S. (France)

    2005-08-15

    The civil and military utilization of nuclear power results in continuously increasing stockpiles of spent fuel and separated plutonium. Since fast breeder reactors are at present not available, the majority of spent fuel discharged from commercial nuclear reactors is intended for direct final disposal or designated for interim storage. An effective form of intermediate plutonium storage is recycling in thermal reactors. Recycling of the recovered plutonium in commercial light water reactors (LWRs) is currently practiced in Belgium, France, Germany, and Switzerland. The number of mixed-oxide (MOX) assemblies reloaded each year in a large variety of reactors demonstrates that plutonium recycling in LWRs has reached industrial maturity. The status of experience gained today at Framatome ANP confirms the reliability of the design codes and the suitability of fuel assembly and core designs. The validation database for increasing exposures of MOX fuel is being continuously expanded. This provides the basis for further extending the discharge exposures of MOX assemblies and for licensing the use of higher plutonium concentrations. Options to support the weapons plutonium reduction programs and for the development of advanced MOX assembly designs are investigated.

  8. Knowledge and abilities catalog for nuclear power plant operators: boiling water reactors

    SciTech Connect (OSTI)

    Not Available

    1986-09-01

    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.

  9. Warm Water Oxidation Verification - Scoping and Stirred Reactor Tests

    SciTech Connect (OSTI)

    Braley, Jenifer C.; Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2011-06-15

    Scoping tests to evaluate the effects of agitation and pH adjustment on simulant sludge agglomeration and uranium metal oxidation at {approx}95 C were performed under Test Instructions(a,b) and as per sections 5.1 and 5.2 of this Test Plan prepared by AREVA. (c) The thermal testing occurred during the week of October 4-9, 2010. The results are reported here. For this testing, two uranium-containing simulant sludge types were evaluated: (1) a full uranium-containing K West (KW) container sludge simulant consisting of nine predominant sludge components; (2) a 50:50 uranium-mole basis mixture of uraninite [U(IV)] and metaschoepite [U(VI)]. This scoping study was conducted in support of the Sludge Treatment Project (STP) Phase 2 technology evaluation for the treatment and packaging of K-Basin sludge. The STP is managed by CH2M Hill Plateau Remediation Company (CHPRC) for the U.S. Department of Energy. Warm water ({approx}95 C) oxidation of sludge, followed by immobilization, has been proposed by AREVA and is one of the alternative flowsheets being considered to convert uranium metal to UO{sub 2} and eliminate H{sub 2} generation during final sludge disposition. Preliminary assessments of warm water oxidation have been conducted, and several issues have been identified that can best be evaluated through laboratory testing. The scoping evaluation documented here was specifically focused on the issue of the potential formation of high strength sludge agglomerates at the proposed 95 C process operating temperature. Prior hydrothermal tests conducted at 185 C produced significant physiochemical changes to genuine sludge, including the formation of monolithic concretions/agglomerates that exhibited shear strengths in excess of 100 kPa (Delegard et al. 2007).

  10. A flooding induced station blackout analysis for a pressurized water reactor using the RISMC toolkit

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Mandelli, Diego; Prescott, Steven; Smith, Curtis; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua; Kinoshita, Robert

    2015-05-17

    In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code calledmore »NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. The impact of power uprate is determined in terms of both core damage probability and safety margins.« less

  11. EIS-0288-S1: Production of Tritium in a Commercial Light Water Reactor (CLWR) Tritium Readiness Supplemental Environmental Impact Statement

    Broader source: Energy.gov [DOE]

    This Supplemental EIS updates the environmental analyses in DOE’s 1999 EIS for the Production of Tritium in a Commercial Light Water Reactor (CLWR EIS). The CLWR EIS addressed the production of tritium in Tennessee Valley Authority reactors in Tennessee using tritium-producing burnable absorber rods.

  12. Fuel assembly for the production of tritium in light water reactors

    DOE Patents [OSTI]

    Cawley, W.E.; Trapp, T.J.

    1983-06-10

    A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

  13. Supercritical Water Reactor (SCWR) - Survey of Materials Research and Development Needs to Assess Viability

    SciTech Connect (OSTI)

    Philip E. MacDonald

    2003-09-01

    Supercritical water-cooled reactors (SCWRs) are among the most promising advanced nuclear systems because of their high thermal efficiency [i.e., about 45% vs. 33% of current light water reactors (LWRs)] and considerable plant simplification. SCWRs achieve this with superior thermodynamic conditions (i.e., high operating pressure and temperature), and by reducing the containment volume and eliminating the need for recirculation and jet pumps, pressurizer, steam generators, steam separators and dryers. The reference SCWR design in the U.S. is a direct cycle, thermal spectrum, light-water-cooled and moderated reactor with an operating pressure of 25 MPa and inlet/outlet coolant temperature of 280/500 °C. The inlet flow splits, partly to a down-comer and partly to a plenum at the top of the reactor pressure vessel to flow downward through the core in special water rods to the inlet plenum. This strategy is employed to provide good moderation at the top of the core, where the coolant density is only about 15-20% that of liquid water. The SCWR uses a power conversion cycle similar to that used in supercritical fossil-fired plants: high- intermediate- and low-pressure turbines are employed with one moisture-separator re-heater and up to eight feedwater heaters. The reference power is 3575 MWt, the net electric power is 1600 MWe and the thermal efficiency is 44.8%. The fuel is low-enriched uranium oxide fuel and the plant is designed primarily for base load operation. The purpose of this report is to survey existing materials for fossil, fission and fusion applications and identify the materials research and development needed to establish the SCWR viabilitya with regard to possible materials of construction. The two most significant materials related factors in going from the current LWR designs to the SCWR are the increase in outlet coolant temperature from 300 to 500 °C and the possible compatibility issues associated with the supercritical water environment. • Reactor pressure vessel • Pumps and piping

  14. Design for an invertible water bottle to facilitate cleaning and promote sustainable water bottle usage

    E-Print Network [OSTI]

    Metlitz, Matthew S

    2014-01-01

    The goal of this thesis is to explore the design of a reusable water bottle that can be inverted to expose the inside. Being able to directly touch the entire inside of the product could facilitate cleaning and consequently ...

  15. Condensate Water Collection for an Institutional Building in Doha, Qatar: An Opportunity for Water Sustainability 

    E-Print Network [OSTI]

    Bryant, J. A.; Ahmed, T.

    2008-01-01

    conditioning system. Internal loads and additional outside ventilation air all generate considerable latent loads on these systems and exacerbate the already difficult moisture control problem. A manifestation of this load is the liquid water condensate...

  16. Cogeneration of Electricity and Potable Water Using The International Reactor Innovative And Secure (IRIS) Design

    SciTech Connect (OSTI)

    Ingersoll, D.T.; Binder, J.L.; Kostin, V.I.; Panov, Y.K.; Polunichev, V.; Ricotti, M.E.; Conti, D.; Alonso, G.

    2004-10-06

    The worldwide demand for potable water has been steadily growing and is projected to accelerate, driven by a continued population growth and industrialization of emerging countries. This growth is reflected in a recent market survey by the World Resources Institute, which shows a doubling in the installed capacity of seawater desalination plants every ten years. The production of desalinated water is energy intensive, requiring approximately 3-6 kWh/m3 of produced desalted water. At current U.S. water use rates, a dedicated 1000 MW power plant for every one million people would be required to meet our water needs with desalted water. Nuclear energy plants are attractive for large scale desalination application. The thermal energy produced in a nuclear plant can provide both electricity and desalted water without the production of greenhouse gases. A particularly attractive option for nuclear desalination is to couple a desalination plant with an advanced, modular, passively safe reactor design. The use of small-to-medium sized nuclear power plants allows for countries with smaller electrical grid needs and infrastructure to add new electrical and water capacity in more appropriate increments and allows countries to consider siting plants at a broader number of distributed locations. To meet these needs, a modified version of the International Reactor Innovative and Secure (IRIS) nuclear power plant design has been developed for the cogeneration of electricity and desalted water. The modular, passively safe features of IRIS make it especially well adapted for this application. Furthermore, several design features of the IRIS reactor will ensure a safe and reliable source of energy and water even for countries with limited nuclear power experience and infrastructure. The IRIS-D design utilizes low-quality steam extracted from the low-pressure turbine to boil seawater in a multi-effect distillation desalination plant. The desalination plant is based on the horizontal tube film evaporation design used successfully with the BN-350 nuclear plant in Aktau, Kazakhstan. Parametric studies have been performed to optimize the balance of plant design. Also, an economic analysis has been performed, which shows that IRIS-D should be able to provide electricity and clean water at highly competitive costs.

  17. Analysis of the magnetic corrosion product deposits on a boiling water reactor cladding

    SciTech Connect (OSTI)

    Orlov, Andrey; Degueldre, Claude; Kaufmann, Wilfried

    2013-01-15

    The buildup of corrosion product deposits (CRUD) on the fuel cladding of the boiling water reactor (BWR) before and after zinc injection has been investigated by applying local experimental analytical techniques. Under the BWR water chemistry conditions, Zn addition together with the presence of Ni and Mn induce the formation of (Zn,Ni,Mn)[Fe{sub 2}O{sub 4}] spinel solid solutions. X-ray absorption spectroscopy (XAS) revealed inversion ratios of cation distribution in spinels deposited from the solid solution. Based on this information, a two-site ferrite spinel solid solution model is proposed. Electron probe microanalysis (EPMA) and extended X-ray absorption fine structure (EXAFS) findings suggest the zinc-rich ferrite spinels formation on BWR fuel cladding mainly at lower pin. - Graphical Abstract: Analysis of spinels in corrosion product deposits on boiling water reactor fuel rod. Combining EPMA and XAFS results: schematic representation of the ferrite spinels in terms of the end members and their extent of inversion. Note that the ferrites are represented as a surface between the normal (upper plane, M[Fe{sub 2}]O{sub 4}) and the inverse (lower plane, Fe[MFe]O{sub 4}). Actual compositions red Black-Small-Square for the specimen at low elevation (810 mm), blue Black-Small-Square for the specimen at mid elevation (1800 mm). The results have an impact on the properties of the CRUD material. Highlights: Black-Right-Pointing-Pointer Buildup of corrosion product deposits on fuel claddings of a boiling water reactor (BWR) are investigated. Black-Right-Pointing-Pointer Under BWR water conditions, Zn addition with Ni and Mn induced formation of (Zn,Ni,Mn)[Fe{sub 2}O{sub 4}]. Black-Right-Pointing-Pointer X-Ray Adsorption Spectroscopy (XAS) revealed inversion of cations in spinel solid solutions. Black-Right-Pointing-Pointer Zinc-rich ferrite spinels are formed on BWR fuel cladding mainly at lower pin elevations.

  18. Advanced fuel assembly characterization capabilities based on gamma tomography at the Halden boiling water reactor

    SciTech Connect (OSTI)

    Holcombe, S.; Eitrheim, K.; Svaerd, S. J.; Hallstadius, L.; Willman, C.

    2012-07-01

    Characterization of individual fuel rods using gamma spectroscopy is a standard part of the Post Irradiation Examinations performed on experimental fuel at the Halden Boiling Water Reactor. However, due to handling and radiological safety concerns, these measurements are presently carried out only at the end of life of the fuel, and not earlier than several days or weeks after its removal from the reactor core. In order to enhance the fuel characterization capabilities at the Halden facilities, a gamma tomography measurement system is now being constructed, capable of characterizing fuel assemblies on a rod-by-rod basis in a more timely and efficient manner. Gamma tomography for measuring nuclear fuel is based on gamma spectroscopy measurements and tomographic reconstruction techniques. The technique, previously demonstrated on irradiated commercial fuel assemblies, is capable of determining rod-by-rod information without the need to dismantle the fuel. The new gamma tomography system will be stationed close to the Halden reactor in order to limit the need for fuel transport, and it will significantly reduce the time required to perform fuel characterization measurements. Furthermore, it will allow rod-by-rod fuel characterization to occur between irradiation cycles, thus allowing for measurement of experimental fuel repeatedly during its irradiation lifetime. The development of the gamma tomography measurement system is a joint project between the Inst. for Energy Technology - OECD Halden Reactor Project, Westinghouse (Sweden), and Uppsala Univ.. (authors)

  19. Modeling of the performance of weapons MOX fuel in light water reactors

    SciTech Connect (OSTI)

    Alvis, J.; Bellanger, P.; Medvedev, P.G.; Peddicord, K.L.; Gellene, G.I.

    1999-05-01

    Both the Russian Federation and the US are pursing mixed uranium-plutonium oxide (MOX) fuel in light water reactors (LWRs) for the disposition of excess plutonium from disassembled nuclear warheads. Fuel performance models are used which describe the behavior of MOX fuel during irradiation under typical power reactor conditions. The objective of this project is to perform the analysis of the thermal, mechanical, and chemical behavior of weapons MOX fuel pins under LWR conditions. If fuel performance analysis indicates potential questions, it then becomes imperative to assess the fuel pin design and the proposed operating strategies to reduce the probability of clad failure and the associated release of radioactive fission products into the primary coolant system. Applying the updated code to anticipated fuel and reactor designs, which would be used for weapons MOX fuel in the US, and analyzing the performance of the WWER-100 fuel for Russian weapons plutonium disposition are addressed in this report. The COMETHE code was found to do an excellent job in predicting fuel central temperatures. Also, despite minor predicted differences in thermo-mechanical behavior of MOX and UO{sub 2} fuels, the preliminary estimate indicated that, during normal reactor operations, these deviations remained within limits foreseen by fuel pin design.

  20. Dose rate estimates from irradiated light-water-reactor fuel assemblies in air

    SciTech Connect (OSTI)

    Lloyd, W.R.; Sheaffer, M.K.; Sutcliffe, W.G.

    1994-01-31

    It is generally considered that irradiated spent fuel is so radioactive (self-protecting) that it can only be moved and processed with specialized equipment and facilities. However, a small, possibly subnational, group acting in secret with no concern for the environment (other than the reduction of signatures) and willing to incur substantial but not lethal radiation doses, could obtain plutonium by stealing and processing irradiated spent fuel that has cooled for several years. In this paper, we estimate the dose rate at various distances and directions from typical pressurized-water reactor (PWR) and boiling-water reactor (BWR) spent-fuel assemblies as a function of cooling time. Our results show that the dose rate is reduced rapidly for the first ten years after exposure in the reactor, and that it is reduced by a factor of {approx}10 (from the one year dose rate) after 15 years. Even for fuel that has cooled for 15 years, a lethal dose (LD50) of 450 rem would be received at 1 m from the center of the fuel assembly after several minutes. However, moving from 1 to 5 m reduces the dose rate by over a factor of 10, and moving from 1 to 10 m reduces the dose rate by about a factor of 50. The dose rates 1 m from the top or bottom of the assembly are considerably less (about 10 and 22%, respectively) than 1 m from the center of the assembly, which is the direction of the maximum dose rate.

  1. Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

    SciTech Connect (OSTI)

    1995-08-01

    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.

  2. Passive containment cooling system with drywell pressure regulation for boiling water reactor

    DOE Patents [OSTI]

    Hill, Paul R. (Tucson, AZ)

    1994-01-01

    A boiling water reactor having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit.

  3. Transactions of the Twenty-First Water Reactor Safety Information Meeting

    SciTech Connect (OSTI)

    Monteleone, S.

    1993-10-01

    This report contains summaries of papers on reactor safety research to be presented at the 21st Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel, Bethesda, Maryland, October 25--27, 1993. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting and are given in the order of their presentation in each session.

  4. Passive containment cooling system with drywell pressure regulation for boiling water reactor

    DOE Patents [OSTI]

    Hill, P.R.

    1994-12-27

    A boiling water reactor is described having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit. 4 figures.

  5. Transactions of the twenty-fifth water reactor safety information meeting

    SciTech Connect (OSTI)

    Monteleone, S.

    1997-09-01

    This report contains summaries of papers on reactor safety research to be presented at the 25th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 20--22, 1997. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion of information exchanged during the course of the meeting, and are given in order of their presentation in each session.

  6. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, 3rd Quarterly Report

    SciTech Connect (OSTI)

    Mac Donald, Philip Elsworth

    2002-06-01

    The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed.

  7. Feasibility Study of Supercritical Light Water Cooled Reactors for Electrical Power Production, 5th Quarterly Report, October - December 2002

    SciTech Connect (OSTI)

    Philip MacDonald; Jacopo Buongiorno; Cliff Davis; J. Stephen Herring; Kevan Weaver; Ron Latanision; Bryce Mitton; Gary Was; Luca Oriani; Mario Carelli; Dmitry Paramonov; Lawrence Conway

    2003-01-01

    The overall objective of this project is to evaluate the feasibility of supercritical light water cooled reactors for electric power production. The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies for the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If no additional moderator is added to the fuel rod lattice, it is possible to attain fast neutron energy spectrum conditions in a supercritical water-cooled reactor (SCWR). This type of core can make use of either fertile or fertile-free fuel and retain a hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity. One can also add moderation and design a thermal spectrum SCWR that can also burn actinides. The project is organized into three tasks:

  8. Microstructural characteristics of PWR [pressurized water reactor] spent fuel relative to its leaching behavior

    SciTech Connect (OSTI)

    Wilson, C.N.

    1986-01-01

    Microstructural, compositional and thermochemical properties of spent nuclear fuel are discussed relative to its potential performance as a high-level waste form under proposed Nevada Nuclear Waste Storage Investigations Project tuff repository conditions. Pressurized water reactor spent fuel specimens with various artificially induced cladding defects were leach tested in deionized water and in a reference tuff groundwater under ambient hot cell air and temperature conditions. Greater fractional actinide release was observed with bare fuel than with clad fuel leached through a cladding defect. Congruent actinide release and preferential release of cesium and technetium were observed in both water types. Selected summary radionuclide release data are presented and correlated to pre- and post-test microstructural characterization data.

  9. Regulatory Concerns on the In-Containment Water Storage System of the Korean Next Generation Reactor

    SciTech Connect (OSTI)

    Ahn, Hyung-Joon; Lee, Jae-Hun; Bang, Young-Seok; Kim, Hho-Jung

    2002-07-15

    The in-containment water storage system (IWSS) is a newly adopted system in the design of the Korean Next Generation Reactor (KNGR). It consists of the in-containment refueling water storage tank, holdup volume tank, and cavity flooding system (CFS). The IWSS has the function of steam condensation and heat sink for the steam release from the pressurizer and provides cooling water to the safety injection system and containment spray system in an accident condition and to the CFS in a severe accident condition. With the progress of the KNGR design, the Korea Institute of Nuclear Safety has been developing Safety and Regulatory Requirements and Guidances for safety review of the KNGR. In this paper, regarding the IWSS of the KNGR, the major contents of the General Safety Criteria, Specific Safety Requirements, Safety Regulatory Guides, and Safety Review Procedures were introduced, and the safety review items that have to be reviewed in-depth from the regulatory viewpoint were also identified.

  10. Examinations of Oxidation and Sulfidation of Grain Boundaries in Alloy 600 Exposed to Simulated Pressurized Water Reactor Primary Water

    SciTech Connect (OSTI)

    Schreiber, Daniel K.; Olszta, Matthew J.; Saxey, David W.; Kruska, Karen; Moore, K. L.; Lozano-Perez, Sergio; Bruemmer, Stephen M.

    2013-06-01

    High-resolution characterizations of intergranular attack in alloy 600 (Ni-17Cr-9Fe) exposed to 325 °C simulated pressurized water reactor (PWR) primary water have been conducted using a combination of scanning electron microscopy, NanoSIMS, analytical transmission electron microscopy and atom probe tomography. The intergranular attack exhibited a two-stage microstructure that consisted of continuous corrosion/oxidation to a depth of ~200 nm from the surface followed by discrete Cr-rich sulfides to a further depth of ~500 nm. The continuous oxidation region contained primarily nanocrystalline MO-structure oxide particles and ended at Ni-rich, Cr-depleted grain boundaries with spaced CrS precipitates. Three-dimensional characterization of the sulfidized region using site-specific atom probe tomography revealed extraordinary grain boundary composition changes, including total depletion of Cr across a several nm wide dealloyed zone as a result of grain boundary migration.

  11. A study of out-of-phase power instabilities in boiling water reactors

    SciTech Connect (OSTI)

    March-Leuba, J.; Blakeman, E.D.

    1988-06-20

    This paper presents a study of the stability of subcritical neutronic modes in boiling water reactors that can result in out-of-phase power oscillations. A mechanism has been identified for this type of oscillation, and LAPUR code has been modified to account for it. Numerical results show that there is a region in the power-flow operating map where an out-or-phase stability mode is likely even if the core-wide mode is stable. 4 refs., 7 figs.

  12. Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

    DOE Patents [OSTI]

    Corletti, Michael M. (New Kensington, PA); Lau, Louis K. (Monroeville, PA); Schulz, Terry L. (Murrysville Boro, PA)

    1993-01-01

    The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

  13. Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

    DOE Patents [OSTI]

    Corletti, M.M.; Lau, L.K.; Schulz, T.L.

    1993-12-14

    The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps. 1 figures.

  14. Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

    SciTech Connect (OSTI)

    Cowell, B.S.; Fisher, S.E.

    1999-02-01

    The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

  15. Swelling in light water reactor internal components: Insights from computational modeling

    SciTech Connect (OSTI)

    Stoller, Roger E.; Barashev, Alexander V.; Golubov, Stanislav I.

    2015-08-01

    A modern cluster dynamics model has been used to investigate the materials and irradiation parameters that control microstructural evolution under the relatively low-temperature exposure conditions that are representative of the operating environment for in-core light water reactor components. The focus is on components fabricated from austenitic stainless steel. The model accounts for the synergistic interaction between radiation-produced vacancies and the helium that is produced by nuclear transmutation reactions. Cavity nucleation rates are shown to be relatively high in this temperature regime (275 to 325°C), but are sensitive to assumptions about the fine scale microstructure produced under low-temperature irradiation. The cavity nucleation rates observed run counter to the expectation that void swelling would not occur under these conditions. This expectation was based on previous research on void swelling in austenitic steels in fast reactors. This misleading impression arose primarily from an absence of relevant data. The results of the computational modeling are generally consistent with recent data obtained by examining ex-service components. However, it has been shown that the sensitivity of the model s predictions of low-temperature swelling behavior to assumptions about the primary damage source term and specification of the mean-field sink strengths is somewhat greater that that observed at higher temperatures. Further assessment of the mathematical model is underway to meet the long-term objective of this research, which is to provide a predictive model of void swelling at relevant lifetime exposures to support extended reactor operations.

  16. Light Water Reactor Sustainability Program Status Report on the Grizzly Code Enhancements

    SciTech Connect (OSTI)

    Stephen R. Novascone; Benjamin W. Spencer; Jason D. Hales

    2013-09-01

    This report summarizes work conducted during fiscal year 2013 to work toward developing a full capability to evaluate fracture contour J-integrals to the Grizzly code. This is a progress report on ongoing work. During the next fiscal year, this capability will be completed, and Grizzly will be capable of evaluating these contour integrals for 3D geometry, including the effects of thermal stress and large deformation. A usable, limited capability has been developed, which is capable of evaluating these integrals on 2D geometry, without considering the effects of material nonlinearity, thermal stress or large deformation. This report presents an overview of the approach used, along with a demonstration of the current capability in Grizzly, including a comparison with an analytical solution.

  17. Light Water Reactor Sustainability Program: survey of models for concrete degradation

    SciTech Connect (OSTI)

    Spencer, Benjamin W; Huang, Hai

    2014-08-01

    Concrete has been used in the construction of nuclear facilities because of two primary properties: its structural strength and its ability to shield radiation. Concrete structures have been known to last for hundreds of years, but they are also known to deteriorate in very short periods of time under adverse conditions. The use of concrete in nuclear facilities for containment and shielding of radiation and radioactive materials has made its performance crucial for the safe operation of the facility. The goal of this report is to review and document the main aging mechanisms of concern for concrete structures in nuclear power plants (NPPs) and the models used in simulations of concrete aging and structural response of degraded concrete structures. This is in preparation for future work to develop and apply models for aging processes and response of aged NPP concrete structures in the Grizzly code. To that end, this report also provides recommendations for developing more robust predictive models for aging effects of performance of concrete.

  18. DOE-NE Light Water Reactor Sustainability Program and EPRI Long...

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

    joint lead Advanced II&C Systems Technologies New Instrumentation and Control and Human System Interfaces and Capabilities (including Advanced II&C Pilot Projects)...

  19. Light Water Reactor Sustainability Program Power Uprate Research and Development Strategy

    SciTech Connect (OSTI)

    Hongbin Zhang

    2011-09-01

    The economic incentives for low-cost electricity generation will continue to drive more plant owners to identify safe and reliable methods to increase the electrical power output of the current nuclear power plant fleet. A power uprate enables a nuclear power plant to increase its electrical output with low cost. However, power uprates brought new challenges to plant owners and operators. These include equipment damage or degraded performance, and unanticipated responses to plant conditions, etc. These problems have arisen mainly from using dated design and safety analysis tools and insufficient understanding of the full implications of the proposed power uprate or from insufficient attention to detail during the design and implementation phase. It is essential to demonstrate that all required safety margins have been properly retained and the existing safety level has been maintained or even increased, with consideration of all the conditions and parameters that have an influence on plant safety. The impact of the power uprate on plant life management for long term operation is also an important issue. Significant capital investments are required to extend the lifetime of an aging nuclear power plant. Power uprates can help the plant owner to recover the investment costs. However, plant aging issues may be aggravated by the power uprate due to plant conditions. More rigorous analyses, inspections and monitoring systems are required.

  20. DOE-NE Light Water Reactor Sustainability Program and EPRI Long-Term

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

    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: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:FinancingPetroleum Based| Department8, 2015 GATEWAY6.1viii ACRONYMS,4-97

  1. Light Water Reactor Sustainability (LWRS) Program - R&D Roadmap for

    Energy Savers [EERE]

    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: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on DeliciousMathematicsEnergyInterested Parties - WAPA PublicLED ADOPTIONto CommercializationLight My Fire...Or9,

  2. Environmentally assisted cracking in light water reactors - annual report, January-December 2001.

    SciTech Connect (OSTI)

    Chopra, O. K.; Chung, H. M.; Clark, R. W.; Gruber, E. E; Hiller, R. W.; Shack, W. J.; Soppet, W. K.; Strain, R. V.; Energy Technology

    2003-06-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from January to December 2001. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of austenitic stainless steels (SSs), (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs, and (c) EAC of Alloy 600. The effects of key material and loading variables, such as strain amplitude, strain rate, temperature, dissolved oxygen (DO) level in water, and material heat treatment, on the fatigue lives of wrought and cast austenitic SSs in air and LWR environments have been evaluated. The mechanism of fatigue crack initiation in austenitic SSs in LWR environments has also been examined. The results indicate that the presence of a surface oxide film or difference in the characteristics of the oxide film has no effect on fatigue crack initiation in austenitic SSs in LWR environments. Slow-strain-rate tensile tests and post-test fractographic analyses were conducted on several model SS alloys irradiated to {approx}2 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) ({approx}3 dpa) in He at 289 C in the Halden reactor. The results were used to determine the influence of alloying and impurity elements on the susceptibility of these steels to IASCC. Corrosion fatigue tests were conducted on nonirradiated austenitic SSs in high-purity water at 289 C to establish the test procedure and conditions that will be used for the tests on irradiated materials. A comprehensive irradiation experiment was initiated to obtain many tensile and disk specimens irradiated under simulated pressurized water reactor conditions at {approx}325 C to 5, 10, 20, and 40 dpa. Crack growth tests were completed on 30% cold-worked Alloy 600 in high-purity water under various environmental and loading conditions. The results are compared with data obtained earlier on several heats of Alloy 600 tested in high-DO water under several heat treatment conditions.

  3. Decommissioning the Romanian Water-Cooled Water-Moderated Research Reactor: New Environmental Perspective on the Management of Radioactive Waste

    SciTech Connect (OSTI)

    Barariu, G.; Giumanca, R. [Romanian Authority for Nuclear Activity (RAAN), Subsidiary of Technology and Engineering for Nuclear Objectives (SITON), 111 Atomistilor St., Bucuresti-Magurele, Ilfov (Romania)

    2006-07-01

    Pre-feasibility and feasibility studies were performed for decommissioning of the water-cooled water-moderated research reactor (WWER) located in Bucharest - Magurele, Romania. Using these studies as a starting point, the preferred safe management strategy for radioactive wastes produced by reactor decommissioning is outlined. The strategy must account for reactor decommissioning, as well as for the rehabilitation of the existing Radioactive Waste Treatment Plant and for the upgrade of the Radioactive Waste Disposal Facility at Baita-Bihor. Furthermore, the final rehabilitation of the laboratories and ecological reconstruction of the grounds need to be provided for, in accordance with national and international regulations. In accordance with IAEA recommendations at the time, the pre-feasibility study proposed three stages of decommissioning. However, since then new ideas have surfaced with regard to decommissioning. Thus, taking into account the current IAEA ideology, the feasibility study proposes that decommissioning of the WWER be done in one stage to an unrestricted clearance level of the reactor building in an Immediate Dismantling option. Different options and the corresponding derived preferred option for waste management are discussed taking into account safety measures, but also considering technical, logistical and economic factors. For this purpose, possible types of waste created during each decommissioning stage are reviewed. An approximate inventory of each type of radioactive waste is presented. The proposed waste management strategy is selected in accordance with the recommended international basic safety standards identified in the previous phase of the project. The existing Radioactive Waste Treatment Plant (RWTP) from the Horia Hulubei Institute for Nuclear Physics and Engineering (IFIN-HH), which has been in service with no significant upgrade since 1974, will need refurbishing due to deterioration, as well as upgrading in order to ensure the plant complies with current safety standards. This plant will also need to be adapted to treat wastes generated by WWER dismantling. The Baita-Bihor National Radioactive Waste Disposal Facility consists of two galleries in an abandoned uranium mine located in the central-western part of the Bihor Mountains in Transylvania. The galleries lie at a depth of 840 m. The facility requires a considerable overhaul. Several steps recommended for the upgrade of the facility are explored. Environmental concerns have lately become a crucial part of the radioactive waste management strategy. As such, all decisions must be made with great regard for land utilization around nuclear objectives. (authors)

  4. Chimney for enhancing flow of coolant water in natural circulation boiling water reactor

    DOE Patents [OSTI]

    Oosterkamp, Willem Jan (Oosterbeek, NL); Marquino, Wayne (San Jose, CA)

    1999-01-05

    A chimney which can be reconfigured or removed during refueling to allow vertical removal of the fuel assemblies. The chimney is designed to be collapsed or dismantled. Collapse or dismantlement of the chimney reduces the volume required for chimney storage during the refueling operation. Alternatively, the chimney has movable parts which allow reconfiguration of its structure. In a first configuration suitable for normal reactor operation, the chimney is radially constricted such that the chimney obstructs vertical removal of the fuel assemblies. In a second configuration suitable for refueling or maintenance of the fuel core, the parts of the chimney which obstruct access to the fuel assemblies are moved radially outward to positions whereat access to the fuel assemblies is not obstructed.

  5. Chimney for enhancing flow of coolant water in natural circulation boiling water reactor

    DOE Patents [OSTI]

    Oosterkamp, W.J.; Marquino, W.

    1999-01-05

    A chimney which can be reconfigured or removed during refueling to allow vertical removal of the fuel assemblies is disclosed. The chimney is designed to be collapsed or dismantled. Collapse or dismantlement of the chimney reduces the volume required for chimney storage during the refueling operation. Alternatively, the chimney has movable parts which allow reconfiguration of its structure. In a first configuration suitable for normal reactor operation, the chimney is radially constricted such that the chimney obstructs vertical removal of the fuel assemblies. In a second configuration suitable for refueling or maintenance of the fuel core, the parts of the chimney which obstruct access to the fuel assemblies are moved radially outward to positions whereas access to the fuel assemblies is not obstructed. 11 figs.

  6. Nuclear Systems Enhanced Performance Program, Maintenance Cycle Extension in Advanced Light Water Reactor Design

    SciTech Connect (OSTI)

    Professor Neill Todreas

    2001-10-01

    A renewed interest in new nuclear power generation in the US has spurred interest in developing advanced reactors with features which will address the public's concerns regarding nuclear generation. However, it is economic performance which will dictate whether any new orders for these plants will materialize. Economic performance is, to a great extent, improved by maximizing the time that the plant is on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Indeed, the strategy for the advanced light water reactor plant IRIS (International Reactor, Innovative and Secure) is to utilize an eight year operating cycle. This report describes a formalized strategy to address, during the design phase, the maintenance-related barriers to an extended operating cycle. The top-level objective of this investigation was to develop a methodology for injecting component and system maintainability issues into the reactor plant design process to overcome these barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the IRIS design. The first step in meeting the top-level objective was to determine the types of operating cycle length barriers that the IRIS design team is likely to face. Evaluation of previously identified regulatory and investment protection surveillance program barriers preventing a candidate operating PWR from achieving an extended (48 month) cycle was conducted in the context of the IRIS design. From this analysis, 54 known IRIS operating cycle length barriers were identified. The resolution methodology was applied to each of these barriers to generate design solution alternatives for consideration in the IRIS design. The methodology developed has been demonstrated to narrow the design space to feasible design solutions which enable a desired operating cycle length, yet is general enough to have broad applicability. Feedback from the IRIS design team indicates that the proposed solutions to the investigated operating cycle length barriers are both feasible and consistent with sound design practice.

  7. Hydordesulfurization of dibenzothiophene using hydrogen generated in situ by the water-gas shift reaction in a trickle bed reactor 

    E-Print Network [OSTI]

    Hook, Bruce David

    1984-01-01

    in situ by the Water ? Gas Shift Reaction in a Trickle Bed Reactor (December 1984) Bruce David Hook, B. S. , Texas AgtM University Chairman of Advisory Committee: Dr. Aydin Akgerman Global depletion of sweet crude oil has required greater use of high.... Hook et al. (1984) have presented a technique for extracting intrinsic kinetics from trickle bed reactors operating at high temperatures and pressures. The Hook et al. technique uses the Soave-Redlich- Kwong equation of state to predict...

  8. A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 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.

    2010-01-01

    discussed, between the Nuclear safety assurance and riskCharges Relating to Nuclear Reactor Safety," 1976, availableof light-water nuclear reactor safety, emphasizing the

  9. Criticality benchmark guide for light-water-reactor fuel in transportation and storage packages

    SciTech Connect (OSTI)

    Lichtenwalter, J.J.; Bowman, S.M.; DeHart, M.D.; Hopper, C.M.

    1997-03-01

    This report is designed as a guide for performing criticality benchmark calculations for light-water-reactor (LWR) fuel applications. The guide provides documentation of 180 criticality experiments with geometries, materials, and neutron interaction characteristics representative of transportation packages containing LWR fuel or uranium oxide pellets or powder. These experiments should benefit the U.S. Nuclear Regulatory Commission (NRC) staff and licensees in validation of computational methods used in LWR fuel storage and transportation concerns. The experiments are classified by key parameters such as enrichment, water/fuel volume, hydrogen-to-fissile ratio (H/X), and lattice pitch. Groups of experiments with common features such as separator plates, shielding walls, and soluble boron are also identified. In addition, a sample validation using these experiments and a statistical analysis of the results are provided. Recommendations for selecting suitable experiments and determination of calculational bias and uncertainty are presented as part of this benchmark guide.

  10. FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL

    SciTech Connect (OSTI)

    Greenspan, Ehud; Todreas, Neil; Taiwo, Temitope

    2009-03-10

    The objective of this DOE NERI program sponsored project was to assess the feasibility of improving the plutonium (Pu) and minor actinide (MA) recycling capabilities of pressurized water reactors (PWRs) by using hydride instead of oxide fuels. There are four general parts to this assessment: 1) Identifying promising hydride fuel assembly designs for recycling Pu and MAs in PWRs 2) Performing a comprehensive systems analysis that compares the fuel cycle characteristics of Pu and MA recycling in PWRs using the promising hydride fuel assembly designs identified in Part 1 versus using oxide fuel assembly designs 3) Conducting a safety analysis to assess the likelihood of licensing hydride fuel assembly designs 4) Assessing the compatibility of hydride fuel with cladding materials and water under typical PWR operating conditions Hydride fuel was found to offer promising transmutation characteristics and is recommended for further examination as a possible preferred option for recycling plutonium in PWRs.

  11. Apparatus for draining lower drywell pool water into suppresion pool in boiling water reactor

    DOE Patents [OSTI]

    Gluntz, Douglas M. (San Jose, CA)

    1996-01-01

    An apparatus which mitigates temperature stratification in the suppression pool water caused by hot water drained into the suppression pool from the lower drywell pool. The outlet of a spillover hole formed in the inner bounding wall of the suppression pool is connected to and in flow communication with one end of piping. The inlet end of the piping is above the water level in the suppression pool. The piping is routed down the vertical downcomer duct and through a hole formed in the thin wall separating the downcomer duct from the suppression pool water. The piping discharge end preferably has an elevation at or near the bottom of the suppression pool and has a location in the horizontal plane which is removed from the point where the piping first emerges on the suppression pool side of the inner bounding wall of the suppression pool. This enables water at the surface of the lower drywell pool to flow into and be discharged at the bottom of the suppression pool.

  12. Crack initiation in smooth fatigue specimens of austenitic stainless steel in light water reactor environments.

    SciTech Connect (OSTI)

    Chopra, O. K.; Smith, J. L.

    1999-04-08

    The fatigue design curves for structural materials specified in Section III of the ASME Boiler and Pressure Vessel Code are based on tests of smooth polished specimens at room temperature in air. The effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves; however, recent test data illustrate the detrimental effects of LWR coolant environments on the fatigue resistance of austenitic stainless steels (SSs). Certain loading and environmental conditions have led to test specimen fatigue lives that are significantly shorter than those obtained in air. Results of fatigue tests that examine the influence of reactor environments on crack initiation and crack growth of austenitic SSs are presented. Block loading was used to mark the fracture surface to determine crack length as a function of fatigue cycles in water environments, Crack lengths were measured by scanning electron microscopy. The mechanism for decreased fatigue life in LWR environments is discussed, and crack growth rates in the smooth fatigue specimens are compared with existing data from studies of crack growth rates.

  13. Study of plutonium disposition using existing GE advanced Boiling Water Reactors

    SciTech Connect (OSTI)

    Not Available

    1994-06-01

    The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the US to dispose of 50 to 100 metric tons of excess of plutonium in a safe and proliferation resistant manner. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing permanent conversion and long-term diversion resistance to this material. The NAS study ``Management and Disposition of Excess Weapons Plutonium identified Light Water Reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a US disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a typical 1155 MWe GE Boiling Water Reactor (BWR) is utilized to convert the plutonium to spent fuel. A companion study of the Advanced BWR has recently been submitted. The MOX core design work that was conducted for the ABWR enabled GE to apply comparable fuel design concepts and consequently achieve full MOX core loading which optimize plutonium throughput for existing BWRs.

  14. Modeling of a Flooding Induced Station Blackout for a Pressurized Water Reactor Using the RISMC Toolkit

    SciTech Connect (OSTI)

    Mandelli, Diego; Prescott, Steven R; Smith, Curtis L; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua J; Kinoshita, Robert A

    2011-07-01

    In the Risk Informed Safety Margin Characterization (RISMC) approach we want to understand not just the frequency of an event like core damage, but how close we are (or are not) to key safety-related events and how might we increase our safety margins. The RISMC Pathway uses the probabilistic margin approach to quantify impacts to reliability and safety by coupling both probabilistic (via stochastic simulation) and mechanistic (via physics models) approaches. This coupling takes place through the interchange of physical parameters and operational or accident scenarios. In this paper we apply the RISMC approach to evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., system activation) and to perform statistical analyses (e.g., run multiple RELAP-7 simulations where sequencing/timing of events have been changed according to a set of stochastic distributions). By using the RISMC toolkit, we can evaluate how power uprate affects the system recovery measures needed to avoid core damage after the PWR lost all available AC power by a tsunami induced flooding. The simulation of the actual flooding is performed by using a smooth particle hydrodynamics code: NEUTRINO.

  15. UBC Social Ecological Economic Development Studies (SEEDS) Student Report Shifting To Sustainable Drinking Water Consumption At UBC: A Social Marketing Plan

    E-Print Network [OSTI]

    Drinking Water Consumption At UBC: A Social Marketing Plan Rosalind Sadowski, Angela Willock University; SHIFTING TO SUSTAINABLE DRINKING WATER CONSUMPTION AT UBC: A SOCIAL MARKETING PLAN ROSALIND SADOWSKI This Initiative Will Succeed............................................... 23 Market Research: What

  16. Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

    SciTech Connect (OSTI)

    Rebak, Raul B.

    2014-12-30

    The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to provide hermetic seal. The replacement of a zirconium alloy using a ferritic material containing chromium and aluminum appears to be the most near term implementation for accident tolerant nuclear fuels.

  17. Chooz A, First Pressurized Water Reactor to be Dismantled in France - 13445

    SciTech Connect (OSTI)

    Boucau, Joseph [Westinghouse Electric Company, 43 rue de l'Industrie, Nivelles (Belgium)] [Westinghouse Electric Company, 43 rue de l'Industrie, Nivelles (Belgium); Mirabella, C. [Westinghouse Electric France, Orsay (France)] [Westinghouse Electric France, Orsay (France); Nilsson, Lennart [Westinghouse Electric Sweden, Vaesteraas (Sweden)] [Westinghouse Electric Sweden, Vaesteraas (Sweden); Kreitman, Paul J. [Westinghouse Electric Company, Lake Bluff, IL 60048 (United States)] [Westinghouse Electric Company, Lake Bluff, IL 60048 (United States); Obert, Estelle [EDF - DPI - CIDEN, Lyon (France)] [EDF - DPI - CIDEN, Lyon (France)

    2013-07-01

    Nine commercial nuclear power plants have been permanently shut down in France to date, of which the Chooz A plant underwent an extensive decommissioning and dismantling program. Chooz Nuclear Power Station is located in the municipality of Chooz, Ardennes region, in the northeast part of France. Chooz B1 and B2 are 1,500 megawatt electric (MWe) pressurized water reactors (PWRs) currently in operation. Chooz A, a 305 MWe PWR implanted in two caves within a hill, began operations in 1967 and closed in 1991, and will now become the first PWR in France to be fully dismantled. EDF CIDEN (Engineering Center for Dismantling and Environment) has awarded Westinghouse a contract for the dismantling of its Chooz A reactor vessel (RV). The project began in January 2010. Westinghouse is leading the project in a consortium with Nuvia France. The project scope includes overall project management, conditioning of the reactor vessel (RV) head, RV and RV internals segmentation, reactor nozzle cutting for lifting the RV out of the pit and seal it afterwards, dismantling of the RV thermal insulation, ALARA (As Low As Reasonably Achievable) forecast to ensure acceptable doses for the personnel, complementary vacuum cleaner to catch the chips during the segmentation work, needs and facilities, waste characterization and packaging, civil work modifications, licensing documentation. The RV and RV internals will be segmented based on the mechanical cutting technology that Westinghouse applied successfully for more than 13 years. The segmentation activities cover the cutting and packaging plan, tooling design and qualification, personnel training and site implementation. Since Chooz A is located inside two caves, the project will involve waste transportation from the reactor cave through long galleries to the waste buffer area. The project will end after the entire dismantling work is completed, and the waste storage is outside the caves and ready to be shipped either to the ANDRA (French National Radioactive Waste Management Agency) waste disposal facilities - (for low-level waste [LLW] and very low-level waste [VLLW], which are considered short lived) - or to the EDF Interim Storage Facility planned to be built on another site - (for low- and intermediate-level waste [LILW], which is considered long lived). The project has started with a detailed conceptual study that determines the step-by-step approach for dismantling the reactor and eventually supplying the packed containers ready for final disposal. All technical reports must be verified and approved by EDF and the French Nuclear Safety Authority before receiving the authorization to start the site work. The detailed conceptual study has been completed to date and equipment design and manufacturing is ongoing. This paper will present the conceptual design of the reactor internals segmentation and packaging process that will be implemented at Chooz A, including the planning, methodology, equipment, waste management, and packaging strategy. (authors)

  18. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics

    SciTech Connect (OSTI)

    Brad Merrill; Melissa Teague; Robert Youngblood; Larry Ott; Kevin Robb; Michael Todosow; Chris Stanek; Mitchell Farmer; Michael Billone; Robert Montgomery; Nicholas Brown; Shannon Bragg-Sitton

    2014-02-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. As a result, continual improvement of technology, including advanced materials and nuclear fuels, remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) initiated an Accident Tolerant Fuel (ATF) Development program. The complex multiphysics behavior of LWR nuclear fuel makes defining specific material or design improvements difficult; as such, establishing qualitative attributes is critical to guide the design and development of fuels and cladding with enhanced accident tolerance. This report summarizes a common set of technical evaluation metrics to aid in the optimization and down selection of candidate designs. As used herein, “metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. Furthermore, this report describes a proposed technical evaluation methodology that can be applied to assess the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed for lead test rod or lead test assembly insertion into a commercial reactor within the desired timeframe (by 2022).

  19. Wastes from selected activities in two light-water reactor fuel cycles

    SciTech Connect (OSTI)

    Palmer, C.R.; Hill, O.F.

    1980-07-01

    This report presents projected volumes and radioactivities of wastes from the production of electrical energy using light-water reactors (LWR). The projections are based upon data developed for a recent environmental impact statement in which the transuranic wastes (i.e., those wastes containing certain long-lived alpha emitters at concentrations of at least 370 becquerels, or 10 nCi, per gram of waste) from fuel cycle activities were characterized. In addition, since the WG.7 assumed that all fuel cycle wastes except mill tailings are placed in a mined geologic repository, the nontransuranic wastes from several activities are included in the projections reported. The LWR fuel cycles considered are the LWR, once-through fuel cycle (Strategy 1), in which spent fuel is packaged in metal canisters and then isolated in geologic formations; and the LWR U/Pu recycle fuel cycle (Strategy 2), wherein spent fuel is reprocessed for recovery and recycle of uranium and plutonium in LWRs. The wastes projected for the two LWR fuel cycles are summarized. The reactor operations and decommissioning were found to dominate the rate of waste generation in each cycle. These activities account for at least 85% of the fuel cycle waste volume (not including head-end wastes) when normalized to per unit electrical energy generated. At 10 years out of reactor, however, spent fuel elements in Strategy 1 represent 98% of the fuel cycle activity but only 4% of the volume. Similarly, the packaged high-level waste, fuel hulls and hardware in Strategy 2 concentrate greater than 95% of the activity in 2% of the waste volume.

  20. Evaluation of a severe accident management strategy for boiling water reactors -- Drywell flooding

    SciTech Connect (OSTI)

    Yu, D.; Xing, L.; Kastenberg, W.E.; Okrent, D. (Univ. of California, Los Angeles, CA (United States). Mechanical, Aerospace, and Nuclear Engineering Dept.)

    1994-05-01

    Flooding of the drywell has been suggested as a strategy to prevent reactor vessel and containment failure in boiling water reactors. To evaluate the candidate strategy, this study considers accident management as a decision problem ( drywell flooding'' versus do nothing'') and develops a decision-oriented framework, namely, the influence diagram approach. This analysis chooses the long-term station blackout sequence for a Mark 1 nuclear power plant (Peach Bottom), and an influence diagram with a single decision node is constructed. The node probabilities in the influence diagram are obtained from US Nuclear Regulatory Commission reports or estimated by probabilistic risk assessment methodology. In assessing potential benefits compared with adverse effects, this analysis uses two consequence measures, i.e., early and late fatalities, as decision criteria. The analysis concludes that even though potential adverse effects exist, such as ex-vessel steam explosions and containment isolation failure, the drywell flooding strategy is preferred to do nothing'' when evaluated in terms of these consequence measures.

  1. Inhalation radiotoxicity of irradiated thorium as a heavy water reactor fuel

    SciTech Connect (OSTI)

    Edwards, G.W.R.; Priest, N.D.; Richardson, R.B. [Atomic Energy of Canada Ltd., Chalk River, Ontario, K0J 1J0 (Canada)

    2013-07-01

    The online refueling capability of Heavy Water Reactors (HWRs), and their good neutron economy, allows a relatively high amount of neutron absorption in breeding materials to occur during normal fuel irradiation. This characteristic makes HWRs uniquely suited to the extraction of energy from thorium. In Canada, the toxicity and radiological protection methods dealing with personnel exposure to natural uranium (NU) spent fuel (SF) are well-established, but the corresponding methods for irradiated thorium fuel are not well known. This study uses software to compare the activity and toxicity of irradiated thorium fuel ('thorium SF') against those of NU. Thorium elements, contained in the inner eight elements of a heterogeneous high-burnup bundle having LEU (Low-enriched uranium) in the outer 35 elements, achieve a similar burnup to NU SF during its residence in a reactor, and the radiotoxicity due to fission products was found to be similar. However, due to the creation of such inhalation hazards as U-232 and Th-228, the radiotoxicity of thorium SF was almost double that of NU SF after sufficient time has passed for the decay of shorter-lived fission products. Current radio-protection methods for NU SF exposure are likely inadequate to estimate the internal dose to personnel to thorium SF, and an analysis of thorium in fecal samples is recommended to assess the internal dose from exposure to this fuel. (authors)

  2. Taking the "waste" out of "wastewater" for human water security and ecosystem sustainability

    E-Print Network [OSTI]

    2012-01-01

    water use) (16). water harvesting depend on engineeringHigher-quality water rainwater-harvesting schemes are Lower-stormwater harvesting b City Regeneration of water, but here

  3. Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

    SciTech Connect (OSTI)

    Woo, H.H.; Lu, S.C.

    1981-09-15

    Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design.

  4. Replacement of outboard main steam isolation valves in a boiling water reactor plant

    SciTech Connect (OSTI)

    Schlereth, J.R.; Pennington, D.

    1996-12-01

    Most Boiling Water Reactor plants utilize wye pattern globe valves for main steam isolation valves for both inboard and outboard isolation. These valves have required a high degree of maintenance attention in order to pass the plant local leakage rate testing (LLRT) requirements at each outage. Northern States Power made a decision in 1993 to replace the outboard valves at it`s Monticello plant with double disc gate valves. The replacement of the outboard valves was completed during the fall outage in 1994. During the spring outage in April of 1996 the first LLRT testing was performed with excellent results. This presentation will address the decision process, time requirements and planning necessary to accomplish the task as well as the performance results and cost effectiveness of replacing these components.

  5. Interfacing systems loss-of-coolant accident in Oconee-1 pressurized water reactor

    SciTech Connect (OSTI)

    Nassersharif, B.

    1984-01-01

    The primary system of a pressurized water reactor (PWR) operates at a relatively high pressure (15.5 MPa, 2250 psia) and consists of piping and components designed to withstand these pressures. The low-pressure-injection system (LPIS) connects to the primary system but possesses low-pressure piping passing outside the containment. Therefore, a potential exists for a loss-of-coolant accident (LOCA) outside the containment and concurrent damage to systems needed to cope with this problem. The emergency core-cooling system (ECCS) is assumed to be available for this event. A set of calculations were performed using the TRAC-PF1 code and a model of the Oconee-1 PWR to investigate the consequences of, and possible operator actions for, such an accident scenario.

  6. Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

    SciTech Connect (OSTI)

    Collins, Emory D.; Delcul, Guillermo D.; Hunt, Rodney D.; Johnson, Jared A.; Spencer, Barry B.

    2014-06-10

    A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

  7. Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

    DOE Patents [OSTI]

    Collins, Emory D; Delcul, Guillermo D; Hunt, Rodney D; Johnson, Jared A; Spencer, Barry B

    2013-11-05

    A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

  8. Considerations for severe-accident management strategies in a pressurized water reactor

    SciTech Connect (OSTI)

    Carbajo, J.J.; Carter, J.C. (IT Corp., Oak Ridge, TN (USA))

    1988-01-01

    This paper presents results of a sensitivity study for potential recovery actions during a station blackout severe accident in a pressurized water reactor (PWR). The accident progression for each of the recovery actions was calculated by a modified version of the severe-accident integrated analysis code MAAP 3.0B. According to MAAP calculations for a station blackout in a PWR, vessel failure can be avoided if the power is recovered between 9500 s (2.6h) and 9800 s (2.7 h) after accident initiation (depending on reflood flow rate) and the core is immediately reflooded. If the power is recovered after 10 000 s (2.8 h), vessel failure cannot be avoided. Finally, there is a time window of 26 h between vessel failure and containment failure for actions that could prevent containment failure.

  9. LWR (Light Water Reactor) power plant simulations using the AD10 and AD100 systems

    SciTech Connect (OSTI)

    Wulff, W.; Cheng, H.S.; Chien, C.J.; Jang, J.Y.; Lin, H.C.; Mallen, A.N.; Wang, S.J. (Brookhaven National Lab., Upton, NY (USA); Institute of Nuclear Energy Research, Lung-Tan (Taiwan); Tawian Power Co., Taipei (Taiwan); Brookhaven National Lab., Upton, NY (USA); Institute of Nuclear Energy Research, Lung-Tan (Taiwan))

    1989-01-01

    Boiling (BWR) and Pressurized (PWR) Water Reactor Power Plants are being simulated at BNL with the AD10 and AD100 Peripheral Processor Systems. The AD10 system has been used for BWR simulations since 1984 for safety analyses, emergency training and optimization studies. BWR simulation capabilities have been implemented recently on the AD100 system and PWR simulation capabilities are currently being developed under the auspices of international cooperation. Modeling and simulation methods are presented with emphasis on the simulation of the Nuclear Steam Supply System. Results are presented for BWR simulation and performance characteristics are compared of the AD10 and AD100 systems. It will be shown that the AD100 simulates two times faster than two AD10 processors operating in parallel and that the computing capacity of one AD100 (with FMU processor) is twice as large as that of two AD10 processors. 9 refs., 5 figs., 1 tab.

  10. Environmentally assisted cracking in light water reactors : semiannual report, July 2000 - December 2000.

    SciTech Connect (OSTI)

    Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.; Strain, R. V.; Energy Technology

    2002-04-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from July 2000 to December 2000. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of primary pressure boundary materials, (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels (SSs), and (c) EAC of Alloys 600 and 690. The fatigue strain-vs.-life data are summarized for the effects of various material, loading, and environmental parameters on the fatigue lives of carbon and low-alloy steels and austenitic SSs. Effects of the reactor coolant environment on the mechanism of fatigue crack initiation are discussed. Two methods for incorporating the effects of LWR coolant environments into the ASME Code fatigue evaluations are presented. Slow-strain-rate tensile tests and posttest fractographic analyses were conducted on several model SS alloys irradiated to {approx}0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) in He at 289 C in the Halden reactor. The results were used to determine the influence of alloying and impurity elements on the susceptibility of these steels to IASCC. A fracture toughness J-R curve test was conducted on a commercial heat of Type 304 SS that was irradiated to {approx}2.0 x 10{sup 21} n {center_dot} cm{sup -2} in the Halden reactor. The results were compared with the data obtained earlier on steels irradiated to 0.3 and 0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) (0.45 and 1.35 dpa). Neutron irradiation at 288 C was found to decrease the fracture toughness of austenitic SSs. Tests were conducted on compact-tension specimens of Alloy 600 under cyclic loading to evaluate the enhancement of crack growth rates in LWR environments. Then, the existing fatigue crack growth data on Alloys 600 and 690 were analyzed to establish the effects of temperature, load ratio, frequency, and stress intensity range on crack growth rates in air.

  11. Final report for the Light Water Breeder Reactor proof-of-breeding analytical support project

    SciTech Connect (OSTI)

    Graczyk, D.G.; Hoh, J.C.; Martino, F.J.; Nelson, R.E.; Osudar, J.; Levitz, N.M.

    1987-05-01

    The technology of breeding /sup 233/U from /sup 232/Th in a light water reactor is being developed and evaluated by the Westinghouse Bettis Atomic Power Laboratory (BAPL) through operation and examination of the Shippingport Light Water Breeder Reactor (LWBR). Bettis is determining the end-of-life (EOL) inventory of fissile uranium in the LWBR core by nondestructive assay of a statistical sample comprising approximately 500 EOL fuel rods. This determination is being made with an irradiated-fuel assay gauge based on neutron interrogation and detection of delayed neutrons from each rod. The EOL fissile inventory will be compared with the beginning-of-life fissile loading of the LWBR to determine the extent of breeding. In support of the BAPL proof-of-breeding (POB) effort, Argonne National Laboratory (ANL) carried out destructive physical, chemical, and radiometric analyses on 17 EOL LWBR fuel rods that were previously assayed with the nondestructive gauge. The ANL work included measurements on the intact rods; shearing of the rods into pre-designated contiguous segments; separate dissolution of each of the more than 150 segments; and analysis of the dissolver solutions to determine each segment's uranium content, uranium isotopic composition, and loading of selected fission products. This report describes the facilities in which this work was carried out, details operations involved in processing each rod, and presents a comprehensive discussion of uncertainties associated with each result of the ANL measurements. Most operations were carried out remotely in shielded cells. Automated equipment and procedures, controlled by a computer system, provided error-free data acquisition and processing, as well as full replication of operations with each rod. Despite difficulties that arose during processing of a few rod segments, the ANL destructive-assay results satisfied the demanding needs of the parent LWBR-POB program.

  12. Study of plutonium disposition using the GE Advanced Boiling Water Reactor (ABWR)

    SciTech Connect (OSTI)

    1994-04-30

    The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the U.S. to disposition 50 to 100 metric tons of excess of plutonium in parallel with a similar program in Russia. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing long-term diversion resistance to this material. The NAS study {open_quotes}Management and Disposition of Excess Weapons Plutonium{close_quotes} identified light water reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a U.S. disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a 1350 MWe GE Advanced Boiling Water Reactor (ABWR) is utilized to convert the plutonium to spent fuel. The ABWR represents the integration of over 30 years of experience gained worldwide in the design, construction and operation of BWRs. It incorporates advanced features to enhance reliability and safety, minimize waste and reduce worker exposure. For example, the core is never uncovered nor is any operator action required for 72 hours after any design basis accident. Phase 1 of this study was documented in a GE report dated May 13, 1993. DOE`s Phase 1 evaluations cited the ABWR as a proven technical approach for the disposition of plutonium. This Phase 2 study addresses specific areas which the DOE authorized as appropriate for more in-depth evaluations. A separate report addresses the findings relative to the use of existing BWRs to achieve the same goal.

  13. ADDITIONAL STRESS AND FRACTURE MECHANICS ANALYSES OF PRESSURIZED WATER REACTOR PRESSURE VESSEL NOZZLES

    SciTech Connect (OSTI)

    Walter, Matthew; Yin, Shengjun; Stevens, Gary; Sommerville, Daniel; Palm, Nathan; Heinecke, Carol

    2012-01-01

    In past years, the authors have undertaken various studies of nozzles in both boiling water reactors (BWRs) and pressurized water reactors (PWRs) located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Those studies described stress and fracture mechanics analyses performed to assess various RPV nozzle geometries, which were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-life (EOL) to require evaluation of embrittlement as part of the RPV analyses associated with pressure-temperature (P-T) limits. In this paper, additional stress and fracture analyses are summarized that were performed for additional PWR nozzles with the following objectives: To expand the population of PWR nozzle configurations evaluated, which was limited in the previous work to just two nozzles (one inlet and one outlet nozzle). To model and understand differences in stress results obtained for an internal pressure load case using a two-dimensional (2-D) axi-symmetric finite element model (FEM) vs. a three-dimensional (3-D) FEM for these PWR nozzles. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated. To investigate the applicability of previously recommended linear elastic fracture mechanics (LEFM) hand solutions for calculating the Mode I stress intensity factor for a postulated nozzle corner crack for pressure loading for these PWR nozzles. These analyses were performed to further expand earlier work completed to support potential revision and refinement of Title 10 to the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G, Fracture Toughness Requirements, and are intended to supplement similar evaluation of nozzles presented at the 2008, 2009, and 2011 Pressure Vessels and Piping (PVP) Conferences. This work is also relevant to the ongoing efforts of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, Working Group on Operating Plant Criteria (WGOPC) efforts to incorporate nozzle fracture mechanics solutions into a revision to ASME B&PV Code, Section XI, Nonmandatory Appendix G.

  14. A domain-specific analysis system for examining nuclear reactor simulation data for light-water and sodium-cooled fast reactors

    E-Print Network [OSTI]

    Billings, Jay Jay; Hull, S Forest; Lingerfelt, Eric J; Wojtowicz, Anna

    2014-01-01

    Building a new generation of fission reactors in the United States presents many technical and regulatory challenges. One important challenge is the need to share and present results from new high-fidelity, high-performance simulations in an easily usable way. Since modern multiscale, multi-physics simulations can generate petabytes of data, they will require the development of new techniques and methods to reduce the data to familiar quantities of interest (e.g., pin powers, temperatures) with a more reasonable resolution and size. Furthermore, some of the results from these simulations may be new quantities for which visualization and analysis techniques are not immediately available in the community and need to be developed. This paper describes a new system for managing high-performance simulation results in a domain-specific way that naturally exposes quantities of interest for light water and sodium-cooled fast reactors. It describes requirements to build such a system and the technical challenges faced...

  15. Improvements of fuel failure detection in boiling water reactors using helium measurements

    SciTech Connect (OSTI)

    Larsson, I.; Sihver, L.; Grundin, A.; Helmersson, J. O.

    2012-07-01

    To certify a continuous and safe operation of a boiling water reactor, careful surveillance of fuel integrity is of high importance. The detection of fuel failures can be performed by off-line gamma spectroscopy of off-gas samples and/or by on-line nuclide specific monitoring of gamma emitting noble gases. To establish the location of a leaking fuel rod, power suppression testing can be used. The accuracy of power suppression testing is dependent on the information of the delay time and the spreading of the released fission gases through the systems before reaching the sampling point. This paper presents a method to improve the accuracy of power suppression testing by determining the delay time and gas spreading profile. To estimate the delay time and examine the spreading of the gas in case of a fuel failure, helium was injected in the feed water system at Forsmark 3 nuclear power plant. The measurements were performed by using a helium detector system based on a mass spectrometer installed in the off-gas system. The helium detection system and the results of the experiment are presented in this paper. (authors)

  16. California’s Water Footprint: recent trends and framework for a sustainable transition

    E-Print Network [OSTI]

    Fulton, Julian

    2015-01-01

    Africa. Barbier, EB. 2004. “Water and Economic Growth. ”Sri Lanka: International Water Management Institute. CDF,California Department of Water Resources. 2015. “California

  17. Weapons-grade plutonium dispositioning. Volume 4. Plutonium dispositioning in light water reactors

    SciTech Connect (OSTI)

    Sterbentz, J.W.; Olsen, C.S.; Sinha, U.P.

    1993-06-01

    This study is in response to a request by the Reactor Panel Subcommittee of the National Academy of Sciences (NAS) Committee on International Security and Arms Control (CISAC) to evaluate the feasibility of using plutonium fuels (without uranium) for disposal in existing conventional or advanced light water reactor (LWR) designs and in low temperature/pressure LWR designs that might be developed for plutonium disposal. Three plutonium-based fuel forms (oxides, aluminum metallics, and carbides) are evaluated for neutronic performance, fabrication technology, and material and compatibility issues. For the carbides, only the fabrication technologies are addressed. Viable plutonium oxide fuels for conventional or advanced LWRs include plutonium-zirconium-calcium oxide (PuO{sub 2}-ZrO{sub 2}-CaO) with the addition of thorium oxide (ThO{sub 2}) or a burnable poison such as erbium oxide (Er{sub 2}O{sub 3}) or europium oxide (Eu{sub 2}O{sub 3}) to achieve acceptable neutronic performance. Thorium will breed fissile uranium that may be unacceptable from a proliferation standpoint. Fabrication of uranium and mixed uranium-plutonium oxide fuels is well established; however, fabrication of plutonium-based oxide fuels will require further development. Viable aluminum-plutonium metallic fuels for a low temperature/pressure LWR include plutonium aluminide in an aluminum matrix (PuAl{sub 4}-Al) with the addition of a burnable poison such as erbium (Er) or europium (Eu). Fabrication of low-enriched plutonium in aluminum-plutonium metallic fuel rods was initially established 30 years ago and will require development to recapture and adapt the technology to meet current environmental and safety regulations. Fabrication of high-enriched uranium plate fuel by the picture-frame process is a well established process, but the use of plutonium would require the process to be upgraded in the United States to conform with current regulations and minimize the waste streams.

  18. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    SciTech Connect (OSTI)

    Chodak, P. III

    1996-05-01

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO{sub 2} assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the {sup 239}Pu and {ge}90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

  19. Development of Advanced Accident Tolerant Fuels for Commercial Light Water Reactors

    SciTech Connect (OSTI)

    Bragg-Sitton, Shannon M.

    2014-03-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. Thanks to efforts by both the U.S. government and private companies, nuclear technologies have advanced over time to optimize economic operations in nuclear utilities while ensuring safety. One of the missions of the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) is to develop nuclear fuels and claddings with enhanced accident tolerance. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, DOE-NE initiated Accident Tolerant Fuel (ATF) development as a primary component of the Fuel Cycle Research & Development (FCRD) Advanced Fuels Campaign (AFC). Prior to the unfortunate events at Fukushima, the emphasis for advanced LWR fuel development was on improving nuclear fuel performance in terms of increased burnup for waste minimization, increased power density for power upgrades, and increased fuel reliability. Fukushima highlighted some undesirable performance characteristics of the standard fuel system during severe accidents, including accelerated hydrogen production under certain circumstances. Thus, fuel system behavior under design basis accident and severe accident conditions became the primary focus for advanced fuels while still striving for improved performance under normal operating conditions to ensure that proposed new fuels will be economically viable. The goal of the ATF development effort is to demonstrate performance with a lead test assembly or lead test rod (LTR) or lead test assembly (LTA) irradiation in a commercial power reactor by 2022. Research and development activities are being conducted at multiple DOE national laboratories, universities and within industry with support from the DOE program. A brief program overview and status are provided.

  20. Development of Advanced Accident Tolerant Fuels for Commercial Light Water Reactors

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Bragg-Sitton, Shannon M.

    2014-03-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. Thanks to efforts by both the U.S. government and private companies, nuclear technologies have advanced over time to optimize economic operations in nuclear utilitiesmore »while ensuring safety. One of the missions of the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) is to develop nuclear fuels and claddings with enhanced accident tolerance. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, DOE-NE initiated Accident Tolerant Fuel (ATF) development as a primary component of the Fuel Cycle Research & Development (FCRD) Advanced Fuels Campaign (AFC). Prior to the unfortunate events at Fukushima, the emphasis for advanced LWR fuel development was on improving nuclear fuel performance in terms of increased burnup for waste minimization, increased power density for power upgrades, and increased fuel reliability. Fukushima highlighted some undesirable performance characteristics of the standard fuel system during severe accidents, including accelerated hydrogen production under certain circumstances. Thus, fuel system behavior under design basis accident and severe accident conditions became the primary focus for advanced fuels while still striving for improved performance under normal operating conditions to ensure that proposed new fuels will be economically viable. The goal of the ATF development effort is to demonstrate performance with a lead test assembly or lead test rod (LTR) or lead test assembly (LTA) irradiation in a commercial power reactor by 2022. Research and development activities are being conducted at multiple DOE national laboratories, universities and within industry with support from the DOE program. A brief program overview and status are provided.« less

  1. MODULAR AND FULL SIZE SIMPLIFIED BOILING WATER REACTOR DESIGN WITH FULLY PASSIVE SAFETY SYSTEMS

    SciTech Connect (OSTI)

    M. Ishii; S. T. Revankar; T. Downar; Y. Xu, H. J. Yoon; D. Tinkler; U. S. Rohatgi

    2003-06-16

    OAK B204 The overall goal of this three-year research project was to develop a new scientific design of a compact modular 200 MWe and a full size 1200 MWe simplified boiling water reactors (SBWR). Specific objectives of this research were: (1) to perform scientific designs of the core neutronics and core thermal-hydraulics for a small capacity and full size simplified boiling water reactor, (2) to develop a passive safety system design, (3) improve and validate safety analysis code, (4) demonstrate experimentally and analytically all design functions of the safety systems for the design basis accidents (DBA) and (5) to develop the final scientific design of both SBWR systems, 200 MWe (SBWR-200) and 1200 MWe (SBWR-1200). The SBWR combines the advantages of design simplicity and completely passive safety systems. These advantages fit well within the objectives of NERI and the Department of Energy's focus on the development of Generation III and IV nuclear power. The 3-year research program was structured around seven tasks. Task 1 was to perform the preliminary thermal-hydraulic design. Task 2 was to perform the core neutronic design analysis. Task 3 was to perform a detailed scaling study and obtain corresponding PUMA conditions from an integral test. Task 4 was to perform integral tests and code evaluation for the DBA. Task 5 was to perform a safety analysis for the DBA. Task 6 was to perform a BWR stability analysis. Task 7 was to perform a final scientific design of the compact modular SBWR-200 and the full size SBWR-1200. A no cost extension for the third year was requested and the request was granted and all the project tasks were completed by April 2003. The design activities in tasks 1, 2, and 3 were completed as planned. The existing thermal-hydraulic information, core physics, and fuel lattice information was collected on the existing design of the simplified boiling water reactor. The thermal-hydraulic design were developed. Based on a detailed integral system scaling analysis, design parameters were obtained and designs of the compact modular 200 MWe SBWR and the full size 1200 MWe SBWR were developed. These reactors are provided with passive safety systems. A new passive vacuum breaker check valve was designed to replace the mechanical vacuum beaker check valve. The new vacuum breaker check valve was based on a hydrostatic head, and was fail safe. The performance of this new valve was evaluated both by the thermal-hydraulic code RELAP5 and by the experiments in a scaled SBWR facility, PUMA. In the core neutronic design a core depletion model was implemented to PARCS code. A lattice design for the SBWR fuel assemblies was performed. Design improvements were made to the neutronics/thermal-hydraulics models of SBWR-200 and SBWR-1200, and design analyses of these reactors were performed. The design base accident analysis and evaluation of all the passive safety systems were completed as scheduled in tasks 4 and 5. Initial conditions for the small break loss of coolant accidents (LOCA) and large break LOCA using REALP5 code were obtained. Small and large break LOCA tests were performed and the data was analyzed. An anticipated transient with scram was simulated using the RELAP5 code for SBWR-200. The transient considered was an accidental closure of the main steam isolation valve (MSIV), which was considered to be the most significant transient. The evaluation of the RELAP5 code against experimental data for SBWR-1200 was completed. In task 6, the instability analysis for the three SBWR designs (SBWR-1200, SBWR-600 and SBWR-200) were simulated for start-up transients and the results were similar. Neither the geysering instability, nor the loop type instability was predicted by RAMONA-4B in the startup simulation following the recommended procedure by GE. The density wave oscillation was not observed at all because the power level used in the simulation was not high enough. A study was made of the potential instabilities by imposing an unrealistically high power ramp in a short time period, as suggested by GE. RAMON

  2. Processing Tritiated Water at the Savannah River Site: A Production-Scale Demonstration of a Palladium Membrane Reactor

    SciTech Connect (OSTI)

    Sessions, Kevin L. [Westinghouse Savannah River Company (United States)

    2005-07-15

    The Palladium Membrane Reactor (PMR) process was installed in the Tritium Facilities at the Savannah River Site to perform a production-scale demonstration for the recovery of tritium from tritiated water adsorbed on molecular sieve (zeolite). Unlike the current recovery process that utilizes magnesium, the PMR offers a means to process tritiated water in a more cost effective and environmentally friendly manner. The design and installation of the large-scale PMR process was part of a collaborative effort between the Savannah River Site and Los Alamos National Laboratory.The PMR process operated at the Savannah River Site between May 2001 and April 2003. During the initial phase of operation the PMR processed thirty-four kilograms of tritiated water from the Princeton Plasma Physics Laboratory. The water was processed in fifteen separate batches to yield approximately 34,400 liters (STP) of hydrogen isotopes. Each batch consisted of round-the-clock operations for approximately nine days. In April 2003 the reactor's palladium-silver membrane ruptured resulting in the shutdown of the PMR process. Reactor performance, process performance and operating experiences have been evaluated and documented. A performance comparison between PMR and current magnesium process is also documented.

  3. Many-Group Cross-Section Adjustment Techniques for Boiling Water Reactor Adaptive Simulation

    SciTech Connect (OSTI)

    Jessee, Matthew Anderson

    2011-01-01

    Computational capability has been developed to adjust multigroup neutron cross sections, including self-shielding correction factors, to improve the fidelity of boiling water reactor (BWR) core modeling and simulation. The method involves propagating multigroup neutron cross-section uncertainties through various BWR computational models to evaluate uncertainties in key core attributes such as core k{sub eff}, nodal power distributions, thermal margins, and in-core detector readings. Uncertainty-based inverse theory methods are then employed to adjust multigroup cross sections to minimize the disagreement between BWR core modeling predictions and observed (i.e., measured) plant data. For this paper, observed plant data are virtually simulated in the form of perturbed three-dimensional nodal power distributions with the perturbations sized to represent actual discrepancies between predictions and real plant data. The major focus of this work is to efficiently propagate multigroup neutron cross-section uncertainty through BWR lattice physics and core simulator calculations. The data adjustment equations are developed using a subspace approach that exploits the ill-conditioning of the multigroup cross-section covariance matrix to minimize computation and storage burden. Tikhonov regularization is also employed to improve the conditioning of the data adjustment equations. Expressions are also provided for posterior covariance matrices of both the multigroup cross-section and core attributes uncertainties.

  4. Qualification Requirements of Guided Ultrasonic Waves for Inspection of Piping in Light Water Reactors

    SciTech Connect (OSTI)

    Meyer, Ryan M.; Ramuhalli, Pradeep; Doctor, Steven R.; Bond, Leonard J.

    2013-08-01

    Guided ultrasonic waves (GUW) are being increasingly used for both NDT and monitoring of piping. GUW offers advantages over many conventional NDE technologies due to the ability to inspect large volumes of piping components without significant removal of thermal insulation or protective layers. In addition, regions rendered inaccessible to more conventional NDE technologies may be more accessible using GUW techniques. For these reasons, utilities are increasingly considering the use of GUWs for performing the inspection of piping components in nuclear power plants. GUW is a rapidly evolving technology and its usage for inspection of nuclear power plant components requires refinement and qualification to ensure it is able to achieve consistent and acceptable levels of performance. This paper will discuss potential requirements for qualification of GUW techniques for the inspection of piping components in light water reactors (LWRs). The Nuclear Regulatory Commission has adopted ASME Boiler and Pressure Vessel Code requirements in Sections V, III, and XI for nondestructive examination methods, fabrication inspections, and pre-service and in-service inspections. A Section V working group has been formed to place the methodology of GUW into the ASME Boiler and Pressure Vessel Code but no requirements for technique, equipment, or personnel exist in the Code at this time.

  5. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics Executive Summary

    SciTech Connect (OSTI)

    Shannon Bragg-Sitton

    2014-02-01

    Research and development (R&D) activities on advanced, higher performance Light Water Reactor (LWR) fuels have been ongoing for the last few years. Following the unfortunate March 2011 events at the Fukushima Nuclear Power Plant in Japan, the R&D shifted toward enhancing the accident tolerance of LWRs. Qualitative attributes for fuels with enhanced accident tolerance, such as improved reaction kinetics with steam resulting in slower hydrogen generation rate, provide guidance for the design and development of fuels and cladding with enhanced accident tolerance. A common set of technical metrics should be established to aid in the optimization and down selection of candidate designs on a more quantitative basis. “Metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. This report describes a proposed technical evaluation methodology that can be applied to evaluate the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed toward qualification.

  6. An Advanced Computational Scheme for the Optimization of 2D Radial Reflectors in Pressurized Water Reactors

    E-Print Network [OSTI]

    Thomas Clerc; Alain Hébert; Hadrien Leroyer; Jean-Philippe Argaud; Bertrand Bouriquet; Agélique Ponçot

    2014-05-12

    This paper presents a computational scheme for the determination of equivalent 2D multi-group heterogeneous reflectors in a Pressurized Water Reactor (PWR). The proposed strategy is to define a full-core calculation consistent with a reference lattice code calculation such as the Method Of Characteristics (MOC) as implemented in APOLLO2 lattice code. The computational scheme presented here relies on the data assimilation module known as "Assimilation de donn\\'{e}es et Aide \\`{a} l'Optimisation (ADAO)" of the SALOME platform developed at \\'{E}lectricit\\'{e} De France (EDF), coupled with the full-core code COCAGNE and with the lattice code APOLLO2. A first validation of the computational scheme is made using the OPTEX reflector model developed at \\'{E}cole Polytechnique de Montr\\'{e}al (EPM). As a result, we obtain 2D multi-group, spatially heterogeneous 2D reflectors, using both diffusion or $\\text{SP}_{\\text{N}}$ operators. We observe important improvements of the power discrepancies distribution over the core when using reflectors computed with the proposed computational scheme, and the $\\text{SP}_{\\text{N}}$ operator enables additional improvements.

  7. Safeguards and security requirements for weapons plutonium disposition in light water reactors

    SciTech Connect (OSTI)

    Thomas, L.L.; Strait, R.S. [Lawrence Livermore National Lab., CA (United States). Fission Energy and Systems Safety Program

    1994-10-01

    This paper explores the issues surrounding the safeguarding of the plutonium disposition process in support of the United States nuclear weapons dismantlement program. It focuses on the disposition of the plutonium by burning mixed oxide fuel in light water reactors (LWR) and addresses physical protection, material control and accountability, personnel security and international safeguards. The S and S system needs to meet the requirements of the DOE Orders, NRC Regulations and international safeguards agreements. Experience has shown that incorporating S and S measures into early facility designs and integrating them into operations provides S and S that is more effective, more economical, and less intrusive. The plutonium disposition safeguards requirements with which the US has the least experience are the implementation of international safeguards on plutonium metal; the large scale commercialization of the mixed oxide fuel fabrication; and the transportation to and loading in the LWRs of fresh mixed oxide fuel. It is in these areas where the effort needs to be concentrated if the US is to develop safeguards and security systems that are effective and efficient.

  8. Evaluation of fission gas release in high-burnup light water reactor fuel rods

    SciTech Connect (OSTI)

    Barner, J.O.; Cunningham, M.E.; Freshley, M.D.; Lanning, D.D. )

    1993-05-01

    Research to define the behavior of Zircaloy-clad light water reactor (LWR) UO[sub 2] fuel irradiated to high burnup levels was conducted as part of the High Burnup Effects Program (HBEP). The HBEP was a 12-yr program that ultimately acquired, characterized, irradiated, and examined after irradiation 82 LWR fuel rods ranging in rod-average fuel burnup from 22 to 69 MWd/kgM with a peak pellet burnup of 83 MWd/kg M. A principal emphasis of the HBEP was to evaluate the effect of high burnup on fission gas release. It was confirmed that fission gas release remained as dependent on design and irradiation history parameters at high burnup levels as at low to moderate burnup levels. One observed high-burnup effect was the development of a burnup-dependent microstructure at the fuel pellet surface when pellet-edge burnup exceeded 65 MWd/kgM. This low-temperature rim region' was characterized by a loss of optically definable grain structure, a high volume of porosity, and diffusion of fission gas from the UO[sub 2] matrix to the porosity. Although the rim region has the potential for enhanced fission gas release, it is concluded that no significant enhancement of rod-average fission gas release at high burnup levels was observed for the examined fuel rods.

  9. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    SciTech Connect (OSTI)

    Krishnan, S.; Bhasin, V.; Mahajan, S.C. [Bhabha Atomic Research Centre, Bombay (India)] [and others

    1997-04-01

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300{degrees}C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered.

  10. Experimental Breeder Reactor-II Primary Tank System Wash Water Workshop

    Broader source: Energy.gov [DOE]

    In 1994 Congress ordered the shutdown of the Experimental Breeder Reactor-II (EBR-II) and a closure project was initiated.

  11. Knowledge and abilities catalog for nuclear power plant operators: Pressurized water reactors. Revision 1

    SciTech Connect (OSTI)

    1995-08-01

    This document provides the basis for the development of content-valid licensing examinations for reactor operators and senior reactor operators. The examinations developed using the PWR catalog will cover those topics listed under Title 10, (ode of Federal Regulations Part 55. The PWR catalog contains approximately 5100 knowledge and ability (K/A) statements for reactor operators and senior reactor operators. The catalog is organized into six major sections: Catalog Organization; Generic Knowledge and Abilities; Plant Systems; Emergency and Abnormal Plant Evolutions; Components and Theory.

  12. Sustainable development through beneficial use of produced water for the oil and gas industry 

    E-Print Network [OSTI]

    Siddiqui, Mustafa Ashique

    2002-01-01

    using desalination membranes. Produced water with up to 45,000 ppm total dissolved solids (TDS) can be treated to agricultural water quality water standards having less than 500 ppm TDS. Finally an economic analysis, including capital and operational...

  13. Investigation of particulate corrosion product transients in the primary coolant of the Winfrith steam generating heavy water reactor

    SciTech Connect (OSTI)

    Means, F.A.; Rodliffe, R.S.; Harding, K.

    1980-03-01

    Equipment for on-line counting and sizing of particles has been used to sample coolant from the primary circuit of a water reactor (the Winfrith steam generating heavy water reactor). The particle size distribution is compared with a determination by electron microscopic examination of a filter sample and is shown to be in good agreement. The technique allows transients in coolant-borne particle concentrations to be sufficiently resolved for analysis in terms of postulated particle deposition and resuspension behavior. The deposition behavior is found to be describable by a first-order rate process with rate constants smaller than those that would be predicted from mass transfer considerations. It is concluded that deposition cannot be limited by mass transfer alone.

  14. California’s Water Footprint: recent trends and framework for a sustainable transition

    E-Print Network [OSTI]

    Fulton, Julian

    2015-01-01

    combined-cycle natural gas power plants, wind turbines, andNatural gas direct use water consumption Power plant thermalgas electricity generation water consumption Power plant

  15. Comparing Simulation Results with Traditional PRA Model on a Boiling Water Reactor Station Blackout Case Study

    SciTech Connect (OSTI)

    Zhegang Ma; Diego Mandelli; Curtis Smith

    2011-07-01

    A previous study used RELAP and RAVEN to conduct a boiling water reactor station black-out (SBO) case study in a simulation based environment to show the capabilities of the risk-informed safety margin characterization methodology. This report compares the RELAP/RAVEN simulation results with traditional PRA model results. The RELAP/RAVEN simulation run results were reviewed for their input parameters and output results. The input parameters for each simulation run include various timing information such as diesel generator or offsite power recovery time, Safety Relief Valve stuck open time, High Pressure Core Injection or Reactor Core Isolation Cooling fail to run time, extended core cooling operation time, depressurization delay time, and firewater injection time. The output results include the maximum fuel clad temperature, the outcome, and the simulation end time. A traditional SBO PRA model in this report contains four event trees that are linked together with the transferring feature in SAPHIRE software. Unlike the usual Level 1 PRA quantification process in which only core damage sequences are quantified, this report quantifies all SBO sequences, whether they are core damage sequences or success (i.e., non core damage) sequences, in order to provide a full comparison with the simulation results. Three different approaches were used to solve event tree top events and quantify the SBO sequences: “W” process flag, default process flag without proper adjustment, and default process flag with adjustment to account for the success branch probabilities. Without post-processing, the first two approaches yield incorrect results with a total conditional probability greater than 1.0. The last approach accounts for the success branch probabilities and provides correct conditional sequence probabilities that are to be used for comparison. To better compare the results from the PRA model and the simulation runs, a simplified SBO event tree was developed with only four top events and eighteen SBO sequences (versus fifty-four SBO sequences in the original SBO model). The estimated SBO sequence conditional probabilities from the original SBO model were integrated to the corresponding sequences in the simplified SBO event tree. These results were then compared with the simulation run results.

  16. Comparison of thorium-based fuels with different fissile components in existing boiling water reactors

    E-Print Network [OSTI]

    Demazière, Christophe

    parameters that are essential for reactor safety, like reactivity coefficients and control rod worths reactors Klara Insulander Björk a,b,*, Valentin Fhager a , Christophe Demazière b a Thor Energy, Sommerrogaten 13e15, NO-0255 Oslo, Norway b Chalmers University of Technology, Department of Nuclear Engineering

  17. Features of temperature control of fuel element cladding for pressurized water nuclear reactor “WWER-1000” while simulating reactor accidents

    SciTech Connect (OSTI)

    Zaytsev, P. A.; Priymak, S. V.; Usachev, V. B.; Oleynikov, P. P.; Soldatkin, D. M. [Scientific Research Institute, Scientific Industrial Association LUCH, Podolsk (Russian Federation)] [Scientific Research Institute, Scientific Industrial Association LUCH, Podolsk (Russian Federation)

    2013-09-11

    During the experiments simulating NPR (nuclear power reactor) accidents with a coolant loss fuel elements behavior in a steam-hydrogen medium was studied at the temperature changed with the rate from 1 to 100K/s within the range of 300÷1500 °C. Indications of the thermocouples fixed on the cladding notably differ from real values of the cladding temperatures in the area of measuring junction due to thermal resistance influence of the transition zones “cladding-junction” and “junction-coolant”. The estimating method of a measurement error was considered which can provide adequate accounting of the influence factors. The method is based on thermal probing of a thermocouple by electric current flashing through thermoelements under the coolant presence or absence, a response time registration and processing, calculation of thermal inertia value for a thermocouple junction. A formula was derived for calculation of methodical error under stationary mode and within the stage of linear increase in temperature, which will determine the conditions for the cladding depressurization. Some variants of the formula application were considered, and the values of methodical errors were established which reached ?5% of maximum value by the final moment of the stage of linear increase in the temperature.

  18. Sustainable Energy Resources for Consumers Webinar on Solar Water Heating Transcript

    Office of Energy Efficiency and Renewable Energy (EERE)

    Video recording transcript of a Webinar on Nov. 16, 2010 about residential solar water heating applications

  19. Two Methods for Converting a Heavy-Water Research Reactor to Use Low-Enriched-Uranium Fuel to Improve Proliferation Resistance After Startup

    E-Print Network [OSTI]

    Kemp, R. Scott

    This article demonstrates the feasibility of converting a heavy-water research reactor from natural to low-enriched uranium in order to slow the production of weapon-usable plutonium, even if the core cannot be physically ...

  20. * Corresponding author -kfingerman@berkeley.edu 1 Integrating Water Sustainability into the Low Carbon Fuel Standard

    E-Print Network [OSTI]

    Kammen, Daniel M.

    sustainability as a task under the Alternative and Renewable Fuel and Vehicle Technology Program (AB 118 Carbon Fuel Standard Kevin Fingerman1* , Daniel Kammen1,2 , and Michael O'Hare2 1 Energy & Resources crops are displaced by energy feedstocks. Often, however, a tradeoff exists between minimizing

  1. Integrated Water Gas Shift Membrane Reactors Utilizing Novel, Non Precious Metal Mixed Matrix Membrane

    SciTech Connect (OSTI)

    Ferraris, John

    2013-09-30

    Nanoparticles of zeolitic imidazolate frameworks and other related hybrid materials were prepared by modifying published synthesis procedures by introducing bases, changing stoichiometric ratios, or adjusting reaction conditions. These materials were stable at temperatures >300 °C and were compatible with the polymer matrices used to prepare mixed- matrix membranes (MMMs). MMMs tested at 300 °C exhibited a >30 fold increase in permeability, compared to those measured at 35 °C, while maintaining H{sub 2}/CO{sub 2} selectivity. Measurements at high pressure (up to 30 atm) and high temperature (up to 300 °C) resulted in an increase in gas flux across the membrane with retention of selectivity. No variations in permeability were observed at high pressures at either 35 or 300 °C. CO{sub 2}-induced plasticization was not observed for Matrimid®, VTEC, and PBI polymers or their MMMs at 30 atm and 300 °C. Membrane surface modification by cross-linking with ethanol diamine resulted in an increase in H{sub 2}/CO{sub 2} selectivity at 35 °C. Spectrometric analysis showed that the cross-linking was effective to temperatures <150 °C. At higher temperatures, the cross-linked membranes exhibit a H{sub 2}/CO{sub 2} selectivity similar to the uncross-linked polymer. Performance of the polybenzimidazole (PBI) hollow fibers prepared at Santa Fe Science and Technology (SFST, Inc.) showed increased flux o to a flat PBI membrane. A water-gas shift reactor has been built and currently being optimized for testing under DOE conditions.

  2. Preliminary design study of small long life boiling water reactor (BWR) with tight lattice thorium nitride fuel

    SciTech Connect (OSTI)

    Trianti, Nuri E-mail: szaki@fi.itba.c.id; Su'ud, Zaki E-mail: szaki@fi.itba.c.id; Arif, Idam E-mail: szaki@fi.itba.c.id; Riyana, EkaSapta

    2014-09-30

    Neutronic performance of small long-life boiling water reactors (BWR) with thorium nitride based fuel has been performed. A recent study conducted on BWR in tight lattice environments (with a lower moderator percentage) produces small power reactor which has some specifications, i.e. 10 years operation time, power density of 19.1 watt/cc and maximum excess reactivity of about 4%. This excess reactivity value is smaller than standard reactivity of conventional BWR. The use of hexagonal geometry on the fuel cell of BWR provides a substantial effect on the criticality of the reactor to obtain a longer operating time. Supported by a tight concept lattice where the volume fraction of the fuel is greater than the moderator and fuel, Thorium Nitride give good results for fuel cell design on small long life BWR. The excess reactivity of the reactor can be reduced with the addition of gadolinium as burnable poisons. Therefore the hexagonal tight lattice fuel cell design of small long life BWR that has a criticality more than 20 years of operating time has been obtained.

  3. Design of a boiling water reactor equilibrium core using thorium-uranium fuel

    SciTech Connect (OSTI)

    Francois, J-L.; Nunez-Carrera, A.; Espinosa-Paredes, G.; Martin-del-Campo, C.

    2004-10-06

    In this paper the design of a Boiling Water Reactor (BWR) equilibrium core using thorium is presented; a heterogeneous blanket-seed core arrangement concept was adopted. The design was developed in three steps: in the first step two different assemblies were designed based on the integrated blanket-seed concept, they are the blanket-dummy assembly and the blanket-seed assembly. The integrated blanketseed concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned-out in a once-through cycle. In the second step, a core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average 235U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the assembly. In the third step an in-house code was developed to evaluate the thorium equilibrium core under transient conditions. A stability analysis was also performed. Regarding the stability analysis, five operational states were analyzed; four of them define the traditional instability region corner of the power-flow map and the fifth one is the operational state for the full power condition. The frequency and the boiling length were calculated for each operational state. The frequency of the analyzed operational states was similar to that reported for BWRs; these are close to the unstable region that occurs due to the density wave oscillation phenomena in some nuclear power plants. Four transient analyses were also performed: manual SCRAM, recirculation pumps trip, main steam isolation valves closure and loss of feed water. The results of these transients are similar to those obtained with the traditional UO2 nuclear fuel.

  4. 2014 DOE Sustainability Awards

    Broader source: Energy.gov [DOE]

    The 2014 U.S. Department of Energy (DOE) Sustainability Awards continues efforts to recognize teams and individuals for their outstanding contributions to the Department’s sustainability mission, including accomplishments in managing pollution, waste, energy, water, and vehicle fleets.

  5. Taking the "waste" out of "wastewater" for human water security and ecosystem sustainability

    E-Print Network [OSTI]

    2012-01-01

    transfers of water or desalination of seawater or brackishrelative to the desalination of seawater. By one estimate,

  6. Campus Sustainability Goals Energy & Climate

    E-Print Network [OSTI]

    Jacobs, Lucia

    Campus Sustainability Goals Energy & Climate By 2014, reduce greenhouse gas emissions to 1990 and water consumption and wastewater production; incorporate sustainable design principles into capital levels. Food & Dining By 2020, increase sustainable food purchases by campus foodservice providers

  7. Thermal hydraulic performance analysis of a small integral pressurized water reactor core

    E-Print Network [OSTI]

    Blair, Stuart R. (Stuart Ryan), 1972-

    2003-01-01

    A thermal hydraulic analysis of the International Reactor Innovative and Secure (IRIS) core has been performed. Thermal margins for steady state and a selection of Loss Of Flow Accidents have been assessed using three ...

  8. Water Gas Shift Reaction with A Single Stage Low Temperature Membrane Reactor

    SciTech Connect (OSTI)

    Ciora, Richard J; Liu, Paul KT

    2013-12-31

    Palladium membrane and Palladium membrane reactor were developed under this project for hydrogen separation and purification for fuel cell applications. A full-scale membrane reactor was designed, constructed and evaluated for the reformate produced from a commercial scale methanol reformer. In addition, the Pd membrane and module developed from this project was successfully evaluated in the field for hydrogen purification for commercial fuel cell applications.

  9. Study of Pu consumption in light water reactors: Evaluation of GE advanced boiling water reactor plants, compilation of Phase 1C task reports

    SciTech Connect (OSTI)

    Not Available

    1994-01-15

    This report summarizes the evaluations conducted during Phase 1C of the Pu Disposition Study have provided further results which reinforce the conclusions reached during Phase 1A & 1B: These conclusions clearly establish the benefits of the fission option and the use of the ABWR as a reliable, proven, well-defined and cost-effective means available to disposition the weapons Pu. This project could be implemented in the near-term at a cost and on a schedule being validated by reactor plants currently under construction in Japan and by cost and schedule history and validated plans for MOX plants in Europe. Evaluations conducted during this phase have established that (1) the MOX fuel is licensable based on existing criteria for new fuel with limited lead fuel rod testing, (2) that the applicable requirements for transport, handling and repository storage can be met, and (3) that all the applicable safeguards criteria can be met.

  10. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production Progress Report for Year 1, Quarter 2 (January - March 2002)

    SciTech Connect (OSTI)

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

    2002-03-01

    The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed.

  11. Design of a proteus lattice representative of a burnt and fresh fuel interface at power conditions in light water reactors

    SciTech Connect (OSTI)

    Hursin, M.; Perret, G. [Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland)

    2012-07-01

    The research program LIFE (Large-scale Irradiated Fuel Experiment) between PSI and Swissnuclear has been started in 2006 to study the interaction between large sets of burnt and fresh fuel pins in conditions representative of power light water reactors. Reactor physics parameters such as flux ratios and reaction rate distributions ({sup 235}U and {sup 238}U fissions and {sup 238}U capture) are calculated to estimate an appropriate arrangement of burnt and fresh fuel pins within the central element of the test zone of the zero-power research reactor PROTEUS. The arrangement should minimize the number of burnt fuel pins to ease fuel handling and reduce costs, whilst guaranteeing that the neutron spectrum in both burnt and fresh fuel regions and at their interface is representative of a large uniform array of burnt and fresh pins in the same moderation conditions. First results are encouraging, showing that the burnt/fresh fuel interface is well represented with a 6 x 6 bundle of burnt pins. The second part of the project involves the use of TSUNAMI, CASMO-4E and DAKOTA to perform parametric and optimization studies on the PROTEUS lattice by varying its pitch (P) and fraction of D{sub 2}O in moderator (F{sub D2O}) to be as representative as possible of a power light water reactor core at hot full power conditions at beginning of cycle (BOC). The parameters P and F{sub D2O} that best represent a PWR at BOC are 1.36 cm and 5% respectively. (authors)

  12. Steam Line Break and Station Blackout Transients for Proliferation-Resistant Hexagonal Tight Lattice Boiling Water Reactor

    SciTech Connect (OSTI)

    Rohatgi, Upendra S. [Brookhaven National Laboratory (United States); Jo, Jae H. [Brookhaven National Laboratory (United States); Chung, Bub Dong [Brookhaven National Laboratory (United States); Takahashi, Hiroshi [Brookhaven National Laboratory (United States); Downar, Thomas J. [Purdue University (United States)

    2004-01-15

    Safety analyses of a proliferation-resistant, economically competitive, high-conversion boiling water reactor (HCBWR) fueled with fissile plutonium and fertile thorium oxide fuel elements, and with passive safety systems, are presented here. The HCBWR developed here is characterized by a very tight lattice with a relatively small water volume fraction in the core that therefore operates with a fast reactor neutron spectrum and a considerably improved neutron economy compared to the current generation of light water reactors. The tight lattice core has a very narrow flow channel with a hydraulic diameter less than half of the regular boiling water reactor (BWR) core and, thus, presents a special challenge to core cooling because of reduced water inventory and high friction in the core. The primary safety concern when reducing the moderator-to-fuel ratio and when using a tightly packed lattice arrangement is to maintain adequate cooling of the core during both normal operation and accident scenarios.In the preliminary HCBWR design, the core is placed in a vessel with a large chimney section, and the vessel is connected to the isolation condenser system (ICS). The vessel is placed in containment with the gravity driven cooling system (GDCS) and passive containment cooling system (PCCS) in a configuration similar to General Electric's simplified BWR (SBWR). The safety systems are similar to those of the SBWR; the ICS and PCCS are scaled with power. An internal recirculation pump is placed in the downcomer to augment the buoyancy head provided by the chimney since the buoyancy provided by the chimney alone could not generate sufficient recirculation in the vessel as the tight lattice configuration results in much larger friction in the core than with the SBWR.The constitutive relationships for RELAP5 are assessed for narrow channels, and as a result the heat transfer package is modified. The modified RELAP5 is used to simulate and analyze two of the most limiting events for a tight pitch lattice core: the station blackout and the main-steam-line-break events. The results of the analyses indicate that the HCBWR system will be safely brought to the shutdown condition for these transients.

  13. Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

    SciTech Connect (OSTI)

    S.T. Revankar; W. Zhou; Gavin Henderson

    2008-07-08

    The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to : 1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, 2) assess the RELAP5 and TRACE computer code against the experimental data, and 3) develop mathematical model and ehat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal- hydraulic codes assessment.

  14. Sustainable Management of Flowback Water during Hydraulic Fracturing of Marcellus Shale for Natural Gas Production

    SciTech Connect (OSTI)

    Vidic, Radisav

    2015-01-24

    This study evaluated the feasibility of using abandoned mine drainage (AMD) as make- up water for the reuse of produced water for hydraulic fracturing. There is an abundance of AMD sources near permitted gas wells as documented in this study that can not only serve as makeup water and reduce the demand on high quality water resources but can also as a source of chemicals to treat produced water prior to reuse. The assessment of AMD availability for this purpose based on proximity and relevant regulations was accompanied by bench- and pilot-scale studies to determine optimal treatment to achieve desired water quality for use in hydraulic fracturing. Sulfate ions that are often present in AMD at elevated levels will react with Ba²? and Sr²? in produced water to form insoluble sulfate compounds. Both membrane microfiltration and gravity separation were evaluated for the removal of solids formed as a result of mixing these two impaired waters. Laboratory studies revealed that neither AMD nor barite formed in solution had significant impact on membrane filtration but that some produced waters contained submicron particles that can cause severe fouling of microfiltration membrane. Coagulation/flocculation was found to be an effective process for the removal of suspended solids and both bench- and pilot-scale studies revealed that optimal process conditions can consistently achieve the turbidity of the finished water below 5 NTU. Adjusting the blending ratio of AMD and produced water can achieve the desired effluent sulfate concentration that can be accurately predicted by chemical thermodynamics. Co-treatment of produced water and AMD will result in elevated levels of naturally occurring radioactive materials (NORM) in the solid waste generated in this process due to radium co-precipitation with barium sulfate. Laboratory studies revealed that the mobility of barite that may form in the subsurface due to the presence of sulfate in the fracturing fluid can be controlled by the addition of appropriate antiscalants.

  15. 3D Simulation of Missing Pellet Surface Defects in Light Water Reactor Fuel Rods

    SciTech Connect (OSTI)

    B.W. Spencer; J.D. Hales; S.R. Novascone; R.L. Williamson

    2012-09-01

    The cladding on light water reactor (LWR) fuel rods provides a stable enclosure for fuel pellets and serves as a first barrier against fission product release. Consequently, it is important to design fuel to prevent cladding failure due to mechanical interactions with fuel pellets. Cladding stresses can be effectively limited by controlling power increase rates. However, it has been shown that local geometric irregularities caused by manufacturing defects known as missing pellet surfaces (MPS) in fuel pellets can lead to elevated cladding stresses that are sufficiently high to cause cladding failure. Accurate modeling of these defects can help prevent these types of failures. Nuclear fuel performance codes commonly use a 1.5D (axisymmetric, axially-stacked, one-dimensional radial) or 2D axisymmetric representation of the fuel rod. To study the effects of MPS defects, results from 1.5D or 2D fuel performance analyses are typically mapped to thermo-mechanical models that consist of a 2D plane-strain slice or a full 3D representation of the geometry of the pellet and clad in the region of the defect. The BISON fuel performance code developed at Idaho National Laboratory employs either a 2D axisymmetric or 3D representation of the full fuel rod. This allows for a computational model of the full fuel rod to include local defects. A 3D thermo-mechanical model is used to simulate the global fuel rod behavior, and includes effects on the thermal and mechanical behavior of the fuel due to accumulation of fission products, fission gas production and release, and the effects of fission gas accumulation on thermal conductivity across the fuel-clad gap. Local defects can be modeled simply by including them in the 3D fuel rod model, without the need for mapping between two separate models. This allows for the complete set of physics used in a fuel performance analysis to be included naturally in the computational representation of the local defect, and for the effects of the local defect to be coupled with the global fuel rod model. This approach for modeling fuel with MPS defects is demonstrated and compared with alternative techniques. The effects of varying parameters of the MPS defect are studied using this technique and presented here.

  16. California’s Water Footprint: recent trends and framework for a sustainable transition

    E-Print Network [OSTI]

    Fulton, Julian

    2015-01-01

    water requirements of bioethanol feedstocks. Further detailsthat increased demand for bioethanol, which has mostly beenwater considerations in bioethanol production, while Scown

  17. Taking the "waste" out of "wastewater" for human water security and ecosystem sustainability

    E-Print Network [OSTI]

    2012-01-01

    Higher-quality water rainwater-harvesting schemes are Lower-Substitution e Rainwater harvesting get, which representsb Melbourne rainwater harvesting a Nation is another locally

  18. California’s Water Footprint: recent trends and framework for a sustainable transition

    E-Print Network [OSTI]

    Fulton, Julian

    2015-01-01

    All Nuclear All Conventional Hydropower All (Macknick et al.large volumes of water for hydropower generation, extractiondetermines the state’s hydropower generation, which provides

  19. Efficient and Sustained Photoelectrochemical Water Oxidation by Cobalt Oxide/Silicon Photoanodes with Nanotextured Interfaces

    E-Print Network [OSTI]

    Javey, Ali

    ,,§ Heinz Frei,,§ and Ian D. Sharp*,,§ Joint Center for Artificial Photosynthesis (JCAP), Lawrence Berkeley the catalyst/semiconductor interface. Artificial photosynthesis, by photoelectrochemical splitting of water

  20. Analysis of a duo-selecting membrane reactor for the water-gas shift

    E-Print Network [OSTI]

    Hardy, AliciA Jillian Jackson, 1978-

    2004-01-01

    The water-gas shift reaction is an exothermic and reversible catalytic process that converts carbon monoxide and water (steam) to hydrogen and carbon dioxide. In regard to energy-related issues, the water-gas shift is part ...

  1. Fatigue crack initiation in carbon and low-alloy steels in light water reactor environments : mechanism and prediction.

    SciTech Connect (OSTI)

    Chopra, O. K.; Shack, W. J.

    1998-01-27

    Section 111 of the ASME Boiler and Pressure Vessel Code specifies fatigue design curves for structural materials. The effects of reactor coolant environments are not explicitly addressed by the Code design curves. Recent test data illustrate potentially significant effects of light water reactor (LWR) coolant environments on the fatigue resistance of carbon and low-alloy steels. Under certain loading and environmental conditions, fatigue lives of test specimens may be shorter than those in air by a factor of {approx}70. The crack initiation and crack growth characteristics of carbon and low-alloy steels in LWR environments are presented. Decreases in fatigue life of these steels in high-dissolved-oxygen water are caused primarily by the effect of environment on growth of short cracks < 100 {micro}m in depth. The material and loading parameters that influence fatigue life in LWR environments are defined. Fatigue life is decreased significantly when five conditions are satisfied simultaneously, viz., applied strain range, service temperature, dissolved oxygen in water, and S content in steel are above a threshold level, and loading strain rate is below a threshold value. Statistical models have been developed for estimating the fatigue life of these steels in LWR environments. The significance of the effect of environment on the current Code design curve is evaluated.

  2. Making Sustainable Energy Choices: Insights on the Energy/Water/Land Nexus

    SciTech Connect (OSTI)

    Not Available

    2014-10-01

    This periodic publication summarizes insights from the body of NREL analysis work. In this issue of Analysis Insights, we examine the implications of our energy choices on water, land use, climate, developmental goals, and other factors. Collectively, NREL's work helps policymakers and investors understand and evaluate energy choices within the complex web of connections, or nexus, between energy, water, and land.

  3. The evaluation of the use of metal alloy fuels in pressurized water reactors. Final report

    SciTech Connect (OSTI)

    Lancaster, D.

    1992-10-26

    The use of metal alloy fuels in a PWR was investigated. It was found that it would be feasible and competitive to design PWRs with metal alloy fuels but that there seemed to be no significant benefits. The new technology would carry with it added economic uncertainty and since no large benefits were found it was determined that metal alloy fuels are not recommended. Initially, a benefit was found for metal alloy fuels but when the oxide core was equally optimized the benefit faded. On review of the optimization of the current generation of ``advanced reactors,`` it became clear that reactor design optimization has been under emphasized. Current ``advanced reactors`` are severely constrained. The AP-600 required the use of a fuel design from the 1970`s. In order to find the best metal alloy fuel design, core optimization became a central effort. This work is ongoing.

  4. ALARA Analysis for Shippingport Pressurized Water Reactor Core 2 Fuel Storage in the Canister Storage Building (CSB)

    E-Print Network [OSTI]

    Lewis, M E

    2000-01-01

    The addition of Shippingport Pressurized Water Reactor (PWR) Core 2 Blanket Fuel Assembly storage in the Canister Storage Building (CSB) will increase the total cumulative CSB personnel exposure from receipt and handling activities. The loaded Shippingport Spent Fuel Canisters (SSFCs) used for the Shippingport fuel have a higher external dose rate. Assuming an MCO handling rate of 170 per year (K East and K West concurrent operation), 24-hr CSB operation, and nominal SSFC loading, all work crew personnel will have a cumulative annual exposure of less than the 1,000 mrem limit.

  5. ALARA Analysis for Shippingport Pressurized Water Reactor Core 2 Fuel Storage in the Canister Storage Building (CSB)

    SciTech Connect (OSTI)

    LEWIS, M.E.

    2000-04-06

    The addition of Shippingport Pressurized Water Reactor (PWR) Core 2 Blanket Fuel Assembly storage in the Canister Storage Building (CSB) will increase the total cumulative CSB personnel exposure from receipt and handling activities. The loaded Shippingport Spent Fuel Canisters (SSFCs) used for the Shippingport fuel have a higher external dose rate. Assuming an MCO handling rate of 170 per year (K East and K West concurrent operation), 24-hr CSB operation, and nominal SSFC loading, all work crew personnel will have a cumulative annual exposure of less than the 1,000 mrem limit.

  6. Removal plan for Shippingport pressurized water reactor core 2 blanket fuel assemblies form T plant to the canister storage building

    SciTech Connect (OSTI)

    Lata

    1996-09-26

    This document presents the current strategy and path forward for removal of the Shippingport Pressurized Water Reactor Core 2 blanket fuel assemblies from their existing storage configuration (wet storage within the T Plant canyon) and transport to the Canister Storage Building (designed and managed by the Spent Nuclear Fuel. Division). The removal plan identifies all processes, equipment, facility interfaces, and documentation (safety, permitting, procedures, etc.) required to facilitate the PWR Core 2 assembly removal (from T Plant), transport (to the Canister storage Building), and storage to the Canister Storage Building. The plan also provides schedules, associated milestones, and cost estimates for all handling activities.

  7. Proceedings of the US Nuclear Regulatory Commission twentieth water reactor safety information meeting; Volume 2, Severe accident research, Thermal hydraulics

    SciTech Connect (OSTI)

    Weiss, A.J. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1993-03-01

    This three-volume report contains papers presented at the Twentieth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 21--23, 1992. The papers describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included 10 different papers presented by researchersfrom CEC, China, Finland, France, Germany, Japan, Spain and Taiwan. Selected papers have been processed separately for inclusion in the Energy Science and Technology Database.

  8. Assessment of Possible Cycle Lengths for Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts

    SciTech Connect (OSTI)

    R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag; Kemal O. Pasamehmetoglu

    2012-04-01

    The tri-isotropic (TRISO) fuel developed for High Temperature reactors is known for its extraordinary fission product retention capabilities [1]. Recently, the possibility of extending the use of TRISO particle fuel to Light Water Reactor (LWR) technology, and perhaps other reactor concepts, has received significant attention [2]. The Deep Burn project [3] currently focuses on once-through burning of transuranic fissile and fissionable isotopes (TRU) in LWRs. The fuel form for this purpose is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the TRISO fuel particle design from high temperature reactor technology, but uses SiC as a matrix material rather than graphite. In addition, FCM fuel may also use a cladding made of a variety of possible material, again including SiC as an admissible choice. The FCM fuel used in the Deep Burn (DB) project showed promising results in terms of fission product retention at high burnup values and during high-temperature transients. In the case of DB applications, the fuel loading within a TRISO particle is constituted entirely of fissile or fissionable isotopes. Consequently, the fuel was shown to be capable of achieving reasonable burnup levels and cycle lengths, especially in the case of mixed cores (with coexisting DB and regular LWR UO2 fuels). In contrast, as shown below, the use of UO2-only FCM fuel in a LWR results in considerably shorter cycle length when compared to current-generation ordinary LWR designs. Indeed, the constraint of limited space availability for heavy metal loading within the TRISO particles of FCM fuel and the constraint of low (i.e., below 20 w/0) 235U enrichment combine to result in shorter cycle lengths compared to ordinary LWRs if typical LWR power densities are also assumed and if typical TRISO particle dimensions and UO2 kernels are specified. The primary focus of this summary is on using TRISO particles with up to 20 w/0 enriched uranium kernels loaded in Pressurized Water Reactor (PWR) assemblies. In addition to consideration of this 'naive' use of TRISO fuel in LWRs, several refined options are briefly examined and others are identified for further consideration including the use of advanced, high density fuel forms and larger kernel diameters and TRISO packing fractions. The combination of 800 {micro}m diameter kernels of 20% enriched UN and 50% TRISO packing fraction yielded reactivity sufficient to achieve comparable burnup to present-day PWR fuel.

  9. Water driven : New Orleans City Hall as a sustainable civic center for 21st century

    E-Print Network [OSTI]

    Sangthong, Pholkrit

    2008-01-01

    The devastating struck of Hurricane Katrina in New Orleans in late summer of 2005 was deadly and immense. The storm destroyed over 170 drinking water facilities and 47 wastewater treatments around the city, and resulted ...

  10. Sustainable water supply: rainwater harvesting for multistoried residential apartments in dhaka, bangladesh 

    E-Print Network [OSTI]

    Sultana, Farzana

    2009-05-15

    implications of supplying non-potable demand nodes with either recycled greywater or rainwater. It is suggested that: ? Commercially available simulation tools can be used to represent gross water flows through a new development and to explore different...

  11. Analytical Study of High Concentration PCB Paint at the Heavy Water Components Test Reactor

    SciTech Connect (OSTI)

    Lowry, N.J.

    1998-10-21

    This report provides results of an analytical study of high concentration PCB paint in a shutdown nuclear test reactor located at the US Department of Energy's Savannah River Site (SRS). The study was designed to obtain data relevant for an evaluation of potential hazards associated with the use of and exposure to such paints.

  12. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Progress Report for Work Through September 2003, 2nd Annual/8th Quarterly Report

    SciTech Connect (OSTI)

    Philip E. MacDonald

    2003-09-01

    The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation-IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% vs. about 33% efficiency for current Light Water Reactors, LWRs) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus the need for recirculation and jet pumps, a pressurizer, steam generators, steam separators and dryers is eliminated. The main mission of the SCWR is generation of low-cost electricity. It is built upon two proven technologies, LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which is also in use around the world.

  13. Applicability of GALE-86 Codes to Integral Pressurized Water Reactor designs

    SciTech Connect (OSTI)

    Geelhood, Kenneth J.; Rishel, Jeremy P.

    2012-06-01

    This report describes work that Pacific Northwest National Laboratory is doing to assist the U.S. Nuclear Regulatory Commission (NRC) Office of New Reactors (NRO) staff in their reviews of applications for nuclear power plants using new reactor core designs. These designs include small integral PWRs (IRIS, mPower, and NuScale reactor designs), HTGRs, (pebble-bed and prismatic-block modular reactor designs) and SFRs (4S and PRISM reactor designs). Under this specific task, PNNL will assist the NRC staff in reviewing the current versions of the GALE codes and identify features and limitations that would need to be modified to accommodate the technical review of iPWR and mPower® license applications and recommend specific changes to the code, NUREG-0017, and associated NRC guidance. This contract is necessary to support the licensing of iPWRs with a near-term focus on the B&W mPower® reactor design. While the focus of this review is on the mPower® reactor design, the review of the code and the scope of recommended changes consider a revision of the GALE codes that would make them universally applicable for other types of integral PWR designs. The results of a detailed comparison between PWR and iPWR designs are reported here. Also included is an investigation of the GALE code and its basis and a determination as to the applicability of each of the bases to an iPWR design. The issues investigated come from a list provided by NRC staff, the results of comparing the PWR and iPWR designs, the parameters identified as having a large impact on the code outputs from a recent sensitivity study and the main bases identified in NUREG-0017. This report will provide a summary of the gaps in the GALE codes as they relate to iPWR designs and for each gap will propose what work could be performed to fill that gap and create a version of GALE that is applicable to integral PWR designs.

  14. Preliminary Feasibility, Design, and Hazard Analysis of a Boiling Water Test Loop Within the Idaho National Laboratory Advanced Test Reactor National Scientific User Facility

    SciTech Connect (OSTI)

    Douglas M. Gerstner

    2009-05-01

    The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The ATR and its support facilities are located at the Idaho National Laboratory (INL). A Boiling Water Test Loop (BWTL) is being designed for one of the irradiation test positions within the. The objective of the new loop will be to simulate boiling water reactor (BWR) conditions to support clad corrosion and related reactor material testing. Further it will accommodate power ramping tests of candidate high burn-up fuels and fuel pins/rods for the commercial BWR utilities. The BWTL will be much like the pressurized water loops already in service in 5 of the 9 “flux traps” (region of enhanced neutron flux) in the ATR. The loop coolant will be isolated from the primary coolant system so that the loop’s temperature, pressure, flow rate, and water chemistry can be independently controlled. This paper presents the proposed general design of the in-core and auxiliary BWTL systems; the preliminary results of the neutronics and thermal hydraulics analyses; and the preliminary hazard analysis for safe normal and transient BWTL and ATR operation.

  15. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

    SciTech Connect (OSTI)

    Not Available

    1994-07-01

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.

  16. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 2: Appendices

    SciTech Connect (OSTI)

    Not Available

    1994-07-01

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.

  17. Advanced nuclear reactor safety analysis: the simulation of a small break loss of coolant accident in the simplified boiling water reactor using RELAP5/MOD3.1.1 

    E-Print Network [OSTI]

    Faust, Christophor Randall

    1995-01-01

    The thermal hydraulic simulation code RELAP5/MOD3.1.1 was utilized to model General Electric's Simplified Boiling Water Reactor plant. The model of the plant was subjected to a small break loss of coolant accident occurring from a guillotine shear...

  18. Methods to Study Intervention Sustainability Using Pre-existing, Community Interventions: Examples from the Water, Sanitation and Hygiene Sector

    E-Print Network [OSTI]

    Arnold, Benjamin Ford

    2009-01-01

    water treatment and ii Contents 3.4A.2 Water qualityfor using household water treatment . . . . . . . . . . . .

  19. Joint Statement of Intent Concerning the Arak Heavy Water Reactor Research

    Energy Savers [EERE]

    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: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on DeliciousMathematicsEnergyInterested Parties - WAPA Public CommentInverted Attic9: JohnofReactor Modernization

  20. Heterogeneous Recycling in Fast Reactors

    SciTech Connect (OSTI)

    Forget, Benoit; Pope, Michael; Piet, Steven J.; Driscoll, Michael

    2012-07-30

    Current sodium fast reactor (SFR) designs have avoided the use of depleted uranium blankets over concerns of creating weapons grade plutonium. While reducing proliferation risks, this restrains the reactor design space considerably. This project will analyze various blanket and transmutation target configurations that could broaden the design space while still addressing the non-proliferation issues. The blanket designs will be assessed based on the transmutation efficiency of key minor actinide (MA) isotopes and also on mitigation of associated proliferation risks. This study will also evaluate SFR core performance under different scenarios in which depleted uranium blankets are modified to include minor actinides with or without moderators (e.g. BeO, MgO, B4C, and hydrides). This will be done in an effort to increase the sustainability of the reactor and increase its power density while still offering a proliferation resistant design with the capability of burning MA waste produced from light water reactors (LWRs). Researchers will also analyze the use of recycled (as opposed to depleted) uranium in the blankets. The various designs will compare MA transmutation efficiency, plutonium breeding characteristics, proliferation risk, shutdown margins and reactivity coefficients with a current reference sodium fast reactor design employing homogeneous recycling. The team will also evaluate the out-of-core accumulation and/or burn-down rates of MAs and plutonium isotopes on a cycle-by-cycle basis. This cycle-by-cycle information will be produced in a format readily usable by the fuel cycle systems analysis code, VISION, for assessment of the sustainability of the deployment scenarios.

  1. Modelling and Management of Sustainable Basin-scale Water Resource Systems (Proceedings of a Boulder Symposium, July 1995). IAHS Publ. no. 231, 1995. 359

    E-Print Network [OSTI]

    Tesfatsion, Leigh

    Modelling and Management of Sustainable Basin-scale Water Resource Systems (Proceedings is based on a cost efficient frontier and is called the generalized flexible least squares method. The objective of this study is to apply the FLS method to hydrological time series. Data from the Green River

  2. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01

    in particular, in boiling water reactors) there is a strongdirect contact boiling water fast reactor (PBWFR)”. In:

  3. Evaluation of anticipatory signal to steam generator pressure control program for 700 MWe Indian pressurized heavy water reactor

    SciTech Connect (OSTI)

    Pahari, S.; Hajela, S.; Rammohan, H. P.; Malhotra, P. K.; Ghadge, S. G.

    2012-07-01

    700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is horizontal channel type reactor with partial boiling at channel outlet. Due to boiling, it has a large volume of vapor present in the primary loops. It has two primary loops connected with the help of pressurizer surge line. The pressurizer has a large capacity and is partly filled by liquid and partly by vapor. Large vapor volume improves compressibility of the system. During turbine trip or load rejection, pressure builds up in Steam Generator (SG). This leads to pressurization of Primary Heat Transport System (PHTS). To control pressurization of SG and PHTS, around 70% of the steam generated in SG is dumped into the condenser by opening Condenser Steam Dump Valves (CSDVs) and rest of the steam is released to the atmosphere by opening Atmospheric Steam Discharge Valves (ASDVs) immediately after sensing the event. This is accomplished by adding anticipatory signal to the output of SG pressure controller. Anticipatory signal is proportional to the thermal power of reactor and the proportionality constant is set so that SG pressure controller's output jacks up to ASDV opening range when operating at 100% FP. To simulate this behavior for 700 MWe IPHWR, Primary and secondary heat transport system is modeled. SG pressure control and other process control program have also been modeled to capture overall plant dynamics. Analysis has been carried out with 3-D neutron kinetics coupled thermal hydraulic computer code ATMIKA.T to evaluate the effect of the anticipatory signal on PHT pressure and over all plant dynamics during turbine trip in 700 MWe IPHWR. This paper brings out the results of the analysis with and without considering anticipatory signal in SG pressure control program during turbine trip. (authors)

  4. Light Water Reactor Sustainability Program Technical Basis Guide Describing How to Perform Safety Margin Configuration Risk Management

    SciTech Connect (OSTI)

    Curtis Smith; James Knudsen; Bentley Harwood

    2013-08-01

    The INL has carried out a demonstration of the RISMC approach for the purpose of configuration risk management. We have shown how improved accuracy and realism can be achieved by simulating changes in risk – as a function of different configurations – in order to determine safety margins as the plant is modified. We described the various technical issues that play a role in these configuration-based calculations with the intent that future applications can take advantage of the analysis benefits while avoiding some of the technical pitfalls that are found for these types of calculations. Specific recommendations have been provided on a variety of topics aimed at improving the safety margin analysis and strengthening the technical basis behind the analysis process.

  5. DOE-NE Light Water Reactor Sustainability Program and EPRI Long-Term Operations Program – Joint Research and Development Plan

    SciTech Connect (OSTI)

    Don Williams

    2014-04-01

    Nuclear power has contributed almost 20% of the total amount of electricity generated in the United States over the past two decades. High capacity factors and low operating costs make nuclear power plants (NPPs) some of the most economical power generators available. Further, nuclear power remains the single largest contributor (nearly 70%) of non-greenhouse gas-emitting electric power generation in the United States. Even when major refurbishments are performed to extend operating life, these plants continue to represent cost-effective, low-carbon assets to the nation’s electrical generation capability.

  6. Light Water Reactor Sustainability Program Operator Performance Metrics for Control Room Modernization: A Practical Guide for Early Design Evaluation

    SciTech Connect (OSTI)

    Ronald Boring; Roger Lew; Thomas Ulrich; Jeffrey Joe

    2014-03-01

    As control rooms are modernized with new digital systems at nuclear power plants, it is necessary to evaluate the operator performance using these systems as part of a verification and validation process. There are no standard, predefined metrics available for assessing what is satisfactory operator interaction with new systems, especially during the early design stages of a new system. This report identifies the process and metrics for evaluating human system interfaces as part of control room modernization. The report includes background information on design and evaluation, a thorough discussion of human performance measures, and a practical example of how the process and metrics have been used as part of a turbine control system upgrade during the formative stages of design. The process and metrics are geared toward generalizability to other applications and serve as a template for utilities undertaking their own control room modernization activities.

  7. DOE-NE Light Water Reactor Sustainability Program and EPRI Long-Term Operations Program – Joint Research and Development Plan

    SciTech Connect (OSTI)

    Don Williams

    2012-04-01

    Nuclear power has contributed almost 20% of the total amount of electricity generated in the United States over the past two decades. High capacity factors and low operating costs make nuclear power plants (NPPs) some of the most economical power generators available. Further, nuclear power remains the single largest contributor (nearly 70%) of non-greenhouse gas-emitting electric power generation in the United States. Even when major refurbishments are performed to extend operating life, these plants continue to represent cost-effective, low-carbon assets to the nation's electrical generation capability.

  8. Light Water Reactor Sustainability (LWRS) Initiative Science-Based R&D to Extend Nuclear Plant Operation

    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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirleyEnergyTher i nAand DOE Safety StandardsLabor SeptemberofDepartmentDuty Vehicle CNG TanksItA

  9. DOE-NE Light Water Reactor Sustainability Program and EPRI Long Term Operation Program Â… Joint Research & Development Plan

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

    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: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:FinancingPetroleum Based| Department8, 2015 GATEWAY6.1viii ACRONYMS,4-97 January27-992-24562

  10. Light Water Reactor Sustainability (LWRS) Initiative Science-Based R&D to Extend Nuclear Plant Operation

    Energy Savers [EERE]

    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: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on DeliciousMathematicsEnergyInterested Parties - WAPA PublicLED ADOPTIONto CommercializationLight My Fire...Or9, 2010

  11. Office of Sustainability 2013-2014 Sustainability Ambassador Student Internship

    E-Print Network [OSTI]

    Tipple, Brett

    Office of Sustainability 2013-2014 Sustainability Ambassador Student Internship Organization: The University of Utah Office of Sustainability's primary objective is to decrease the University's environmental of sustainability. We work with the many campus organizations on topics that include energy, food, water, grounds

  12. Blending Cr2O3 into a NiO-Ni electrocatalyst for sustained water splitting

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Gong, Ming; Zhou, Wu; Kenney, Michael James; Kapusta, Rich; Cowley, Sam; Wu, Yingpeng; Lu, Bingan; Lin, Meng -Chang; Wang, Di -Yan; Yang, Jiang; et al

    2015-08-24

    The rising H2 economy demands active and durable electrocatalysts based on low-cost, earth-abundant materials for water electrolysis/photolysis. Here we report nanoscale Ni metal cores over-coated by a Cr2O3-blended NiO layer synthesized on metallic foam substrates. The Ni@NiO/Cr2O3 triphase material exhibits superior activity and stability similar to Pt for the hydrogen-evolution reaction in basic solutions. The chemically stable Cr2O3 is crucial for preventing oxidation of the Ni core, maintaining abundant NiO/Ni interfaces as catalytically active sites in the heterostructure and thus imparting high stability to the hydrogen-evolution catalyst. The highly active and stable electrocatalyst enables an alkaline electrolyzer operating at 20more »mA cm–2 at a voltage lower than 1.5 V, lasting longer than 3 weeks without decay. Thus, the non-precious metal catalysts afford a high efficiency of about 15 % for light-driven water splitting using GaAs solar cells.« less

  13. Boiling water reactor fuel behavior at burnup of 26 GWd/tonne U under reactivity-initiated accident conditions

    SciTech Connect (OSTI)

    Nakamura, Takehiko; Yoshinaga, Makio . Dept. of Reactor Safety Research); Sobajima, Makoto ); Ishijima, Kiyomi; Fujishiro, Toshio . Dept. of Reactor Safety Research)

    1994-10-01

    Irradiated boiling water reactor (BWR) fuel behavior under reactivity-initiated accident (RIA) conditions was investigated in the Nuclear Safety Research Reactor (NSRR) of the Japan Atomic Energy Research Institute. Short test fuel rods, refabricated from a commercial 7 x 7 type BWR fuel rod at a burnup of 26 GWd/ tonne U, were pulse irradiated in the NSRR under simulated cooled startup RIA conditions of the BWRs. Thermal energy from 230 J/g fuel (55 cal/g fuel) to 410 J/g fuel (98 cal/g fuel) was promptly subjected to the test fuel rods by pulse irradiation within [approximately] 10 ms. The peak fuel enthalpies are believed to be the same as the prompt energy depositions. The test fuel rods demonstrated characteristic behavior of the irradiated fuel rods under the accident conditions, such as enhanced pellet cladding mechanical interaction (PCMI) and fission gas release. However, all the fuel rods survived the accident conditions with considerable margins. Simulations by the FRAP-T6 code and fresh fuel rod tests under the same RIA conditions highlighted the burnup effects on the accident fuel performance. The tests and the simulation suggested that the BWR fuel would possibly fail by a cladding burst due to fission gas release during the cladding temperature escalation rather than the PCMI under the cold startup RIA conditions of a severe power burst.

  14. A global approach of the representativity concept: Application on a high-conversion light water reactor MOX lattice case

    SciTech Connect (OSTI)

    Santos, N. D.; Blaise, P.; Santamarina, A.

    2013-07-01

    The development of new types of reactor and the increase in the safety specifications and requirements induce an enhancement in both nuclear data knowledge and a better understanding of the neutronic properties of the new systems. This enhancement is made possible using ad hoc critical mock-up experiments. The main difficulty is to design these experiments in order to obtain the most valuable information. Its quantification is usually made by using representativity and transposition concepts. These theories enable to extract some information about a quantity of interest (an integral parameter) on a configuration, but generally a posteriori. This paper presents a more global approach of this theory, with the idea of optimizing the representativity of a new experiment, and its transposition a priori, based on a multiparametric approach. Using a quadratic sum, we show the possibility to define a global representativity which permits to take into account several quantities of interest at the same time. The maximization of this factor gives information about all quantities of interest. An optimization method of this value in relation to technological parameters (over-clad diameter, atom concentration) is illustrated on a high-conversion light water reactor MOX lattice case. This example tackles the problematic of plutonium experiment for the plutonium aging and a solution through the optimization of both the over-clad and the plutonium content. (authors)

  15. Development of an internally cooled annular fuel bundle for pressurized heavy water reactors

    SciTech Connect (OSTI)

    Hamilton, H.; Armstrong, J.; Kittmer, A.; Zhuchkova, A.; Xu, R.; Hyland, B.; King, M.; Nava-Dominguez, A.; Livingstone, S.; Bergeron, A. [Atomic Energy of Canada, Ltd., Chalk River Laboratories, Chalk River, ON (Canada)

    2013-07-01

    A number of preliminary studies have been conducted at Atomic Energy of Canada Limited to explore the potential of using internally cooled annular fuel (ICAF) in CANDU reactors including finite element thermo-mechanical modelling, reactor physics, thermal hydraulics, fabrication and mechanical design. The most compelling argument for this design compared to the conventional solid-rod design is the significant reduction in maximum fuel temperature for equivalent LERs (linear element ratings). This feature presents the potential for power up-rating or higher burnup and a decreased defect probability due to in-core power increases. The thermal-mechanical evaluation confirmed the significant reduction in maximum fuel temperatures for ICAF fuel compared to solid-rod fuel for equivalent LER. The maximum fuel temperature increase as a function of LER increase is also significantly less for ICAF fuel. As a result, the sheath stress induced by an equivalent power increase is approximately six times less for ICAF fuel than solid-rod fuel. This suggests that the power-increase thresholds to failure (due to stress-corrosion cracking) for ICAF fuel should be well above those for solid-rod fuel, providing improvement in operation flexibility and safety.

  16. An Estimate of the Cost of Electricity from Light Water Reactors and Fossil Plants with Carbon Capture and Sequestration

    SciTech Connect (OSTI)

    Simon, A J

    2009-08-21

    As envisioned in this report, LIFE technology lends itself to large, centralized, baseload (or 'always on') electrical generation. Should LIFE plants be built, they will have to compete in the electricity market with other generation technologies. We consider the economics of technologies with similar operating characteristics: significant economies of scale, limited capacity for turndown, zero dependence on intermittent resources and ability to meet environmental constraints. The five generation technologies examined here are: (1) Light Water Reactors (LWR); (2) Coal; (3) Coal with Carbon Capture and Sequestration (CCS); (4) Natural Gas; and (5) Natural Gas with Carbon Capture and Sequestration. We use MIT's cost estimation methodology (Du and Parsons, 2009) to determine the cost of electricity at which each of these technologies is viable.

  17. Silicon carbide performance as cladding for advanced uranium and thorium fuels for light water reactors

    E-Print Network [OSTI]

    Sukjai, Yanin

    2014-01-01

    There has been an ongoing interest in replacing the fuel cladding zirconium-based alloys by other materials to reduce if not eliminate the autocatalytic and exothermic chemical reaction with water and steam at above 1,200 ...

  18. Sustainable Scientists

    E-Print Network [OSTI]

    Mills, Evan

    2009-01-01

    Exascale for Energy, Ecological Sustainability and Globalrating energy efficiency and other sustainability features,of sustainability often do so in highly energy- intensive

  19. Study of Cost Effective Large Advanced Pressurized Water Reactors that Employ Passive Safety Features

    SciTech Connect (OSTI)

    Winters, J. W.; Corletti, M. M.; Hayashi, Y.

    2003-11-12

    A report of DOE sponsored portions of AP1000 Design Certification effort. On December 16, 1999, The United States Nuclear Regulatory Commission issued Design Certification of the AP600 standard nuclear reactor design. This culminated an 8-year review of the AP600 design, safety analysis and probabilistic risk assessment. The AP600 is a 600 MWe reactor that utilizes passive safety features that, once actuated, depend only on natural forces such as gravity and natural circulation to perform all required safety functions. These passive safety systems result in increased plant safety and have also significantly simplified plant systems and equipment, resulting in simplified plant operation and maintenance. The AP600 meets NRC deterministic safety criteria and probabilistic risk criteria with large margins. A summary comparison of key passive safety system design features is provided in Table 1. These key features are discussed due to their importance in affecting the key thermal-hydraulic phenomenon exhibited by the passive safety systems in critical areas. The scope of some of the design changes to the AP600 is described. These changes are the ones that are important in evaluating the passive plant design features embodied in the certified AP600 standard plant design. These design changes are incorporated into the AP1000 standard plant design that Westinghouse is certifying under 10 CFR Part 52. In conclusion, this report describes the results of the representative design certification activities that were partially supported by the Nuclear Energy Research Initiative. These activities are unique to AP1000, but are representative of research activities that must be driven to conclusion to realize successful licensing of the next generation of nuclear power plants in the United States.

  20. Decommissioning of the secondary containment of the steam generating heavy water reactor at UKAEA-Winfrith

    SciTech Connect (OSTI)

    Miller, Keith; Cornell, Rowland; Parkinson, Steve; McIntyre, Kevin; Staples, Andy

    2007-07-01

    Available in abstract form only. Full text of publication follows: The Winfrith SGHWR was a prototype nuclear power plant operated for 23 years by the United Kingdom Atomic Energy Authority (UKAEA) until 1990 when it was shut down permanently. The current Stage 1 decommissioning contract is part of a multi-stage strategy. It involves the removal of all the ancillary plant and equipment in the secondary containment and non-containment areas ahead of a series of contracts for the decommissioning of the primary containment, the reactor core and demolition of the building and all remaining facilities. As an outcome of a competitive tending process, the Stage 1 decommissioning contract was awarded to NUKEM with operations commencing in April 2005. The decommissioning processes involved with these plant items will be described with some emphasis of the establishment of multiple work-fronts for the production, recovery, treatment and disposal of mainly tritium-contaminated waste arising from its contact with the direct cycle reactor coolant. The means of size reduction of a variety of large, heavy and complex items of plant made from a range of materials will also be described with some emphasis on the control of fumes during hot cutting operations and establishing effective containments within a larger secondary containment structure. Disposal of these wastes in a timely and cost-effective manner is a major challenge facing the decommissioning team and has required the development of a highly efficient means of packing the resultant materials into mainly one-third height ISO containers for disposal as LLW. Details of the quantities of LLW and exempt wastes handled during this process will be given with a commentary about the difficulty in segregating these two waste streams efficiently. (authors)

  1. Thermal neutron steady-state spectra in light water reactor fuel assemblies poisoned with various non-1/v absorbers of different concentrations

    SciTech Connect (OSTI)

    Swaminathan, K.; Chandra, S.; Jha, R.C.; Tewari, S.P. )

    1991-07-01

    This paper reports on the thermal neutron scattering kernel that explicitly incorporates the presence of chemical binding energy and the collective oscillations in the dynamics of water, the steady-state thermal neutron spectra in light water reactor fuel assemblies poisoned with non-1/v absorbers, such as cadmium, samarium, erbium, and gadolinium, in various concentrations have been computed at 298 K. The calculated spectra are in reasonable agreement with the corresponding experimental spectra for realistic source terms.

  2. Review of industry efforts to manage pressurized water reactor feedwater nozzle, piping, and feedring cracking and wall thinning

    SciTech Connect (OSTI)

    Shah, V.N.; Ware, A.G.; Porter, A.M.

    1997-03-01

    This report presents a review of nuclear industry efforts to manage thermal fatigue, flow-accelerated corrosion, and water hammer damage to pressurized water reactor (PWR) feedwater nozzles, piping, and feedrings. The review includes an evaluation of design modifications, operating procedure changes, augmented inspection and monitoring programs, and mitigation, repair and replacement activities. Four actions were taken: (a) review of field experience to identify trends of operating events, (b) review of technical literature, (c) visits to PWR plants and a PWR vendor, and (d) solicitation of information from 8 other countries. Assessment of field experience is that licensees have apparently taken sufficient action to minimize feedwater nozzle cracking caused by thermal fatigue and wall thinning of J-tubes and feedwater piping. Specific industry actions to minimize the wall-thinning in feedrings and thermal sleeves were not found, but visual inspection and necessary repairs are being performed. Assessment of field experience indicates that licensees have taken sufficient action to minimize steam generator water hammer in both top-feed and preheat steam generators. Industry efforts to minimize multiple check valve failures that have allowed backflow of steam from a steam generator and have played a major role in several steam generator water hammer events were not evaluated. A major finding of this review is that analysis, inspection, monitoring, mitigation, and replacement techniques have been developed for managing thermal fatigue and flow-accelerated corrosion damage to feedwater nozzles, piping, and feedrings. Adequate training and appropriate applications of these techniques would ensure effective management of this damage.

  3. Decommissioning of the secondary containment of the steam generating heavy water reactor at UKAEA Winfrith

    SciTech Connect (OSTI)

    Miller, K.D.; Cornell, R.M.; Parkinson, S.J.; McIntyre, K.; Staples, A.

    2007-07-01

    The Winfrith SGHWR was a prototype nuclear power plant operated for 23 years by the United Kingdom Atomic Energy Authority (UKAEA) until 1990 when it was shut down permanently. The current Stage 1 decommissioning contract is part of a multi-stage strategy. It involves the removal of all the ancillary plant and equipment in the secondary containment and non-containment areas ahead of a series of contracts for the decommissioning of the primary containment, the reactor core and demolition of the building and ail remaining facilities. As an outcome of a competitive tending process, the Stage 1 decommissioning contract was awarded to NUKEM with operations commencing in April 2005. The decommissioning processes involved with these plant items will be described with some emphasis of the establishment of multiple work-fronts for the production, recovery, treatment and disposal of mainly tritium-contaminated waste arising from its contact with the direct cycle reactor coolant. The means of size reduction of a variety of large, heavy and complex items of plant made from a range of materials will also be described with some emphasis on the control of fumes during hot cutting operations and establishing effective containments within a larger secondary containment structure. Disposal of these wastes in a timely and cost-effective manner is a major challenge facing the decommissioning team and has required the development of a highly efficient means of packing the resultant materials into mainly one-third height IS0 containers for disposal as LLW. Details of the quantities of LLW and exempt wastes handled during this process will be given with a commentary about the difficulty in segregating these two waste streams efficiently. The paper sets out to demonstrate the considerable progress that has been made with these challenging decommissioning operations at the SGHWR plant and to highlight some of the techniques and processes that have contributed to the overall success of the process. The overall management and control of safety during ail aspects of this challenging contract are major features of the paper, greatly assisted by the adoption from the outset of a non-adversarial team working approach between client and contractor. This has greatly assisted in developing safe and cost-effective solutions to a number of problems that have arisen during the programme, demonstrating the worth of adopting this co-operative approach for mutual benefit. (authors)

  4. Sustainable Development

    E-Print Network [OSTI]

    Brierley, Andrew

    150 Sustainable Development Sustainable Development Degree options BSc or MA (Single Honours Degree) Sustainable Development Contributing Schools Biology; Chemistry; Computer Science; Geography & Geosciences within Sustainable Development, then you should apply for the MA degree; if you are most interested

  5. Bottom head to shell junction assembly for a boiling water nuclear reactor

    DOE Patents [OSTI]

    Fife, A.B.; Ballas, G.J.

    1998-02-24

    A bottom head to shell junction assembly which, in one embodiment, includes an annular forging having an integrally formed pump deck and shroud support is described. In the one embodiment, the annular forging also includes a top, cylindrical shaped end configured to be welded to one end of the pressure vessel cylindrical shell and a bottom, conical shaped end configured to be welded to the disk shaped bottom head. Reactor internal pump nozzles also are integrally formed in the annular forging. The nozzles do not include any internal or external projections. Stubs are formed in each nozzle opening to facilitate welding a pump housing to the forging. Also, an upper portion of each nozzle opening is configured to receive a portion of a diffuser coupled to a pump shaft which extends through the nozzle opening. Diffuser openings are formed in the integral pump deck to provide additional support for the pump impellers. The diffuser opening is sized so that a pump impeller can extend at least partially therethrough. The pump impeller is connected to the pump shaft which extends through the nozzle opening. 5 figs.

  6. Bottom head to shell junction assembly for a boiling water nuclear reactor

    DOE Patents [OSTI]

    Fife, Alex Blair (San Jose, CA); Ballas, Gary J. (San Jose, CA)

    1998-01-01

    A bottom head to shell junction assembly which, in one embodiment, includes an annular forging having an integrally formed pump deck and shroud support is described. In the one embodiment, the annular forging also includes a top, cylindrical shaped end configured to be welded to one end of the pressure vessel cylindrical shell and a bottom, conical shaped end configured to be welded to the disk shaped bottom head. Reactor internal pump nozzles also are integrally formed in the annular forging. The nozzles do not include any internal or external projections. Stubs are formed in each nozzle opening to facilitate welding a pump housing to the forging. Also, an upper portion of each nozzle opening is configured to receive a portion of a diffuser coupled to a pump shaft which extends through the nozzle opening. Diffuser openings are formed in the integral pump deck to provide additional support for the pump impellers. The diffuser opening is sized so that a pump impeller can extend at least partially therethrough. The pump impeller is connected to the pump shaft which extends through the nozzle opening.

  7. Investigations on optimization of accident management measures following a station blackout accident in a VVER-1000 pressurized water reactor

    SciTech Connect (OSTI)

    Tusheva, P.; Schaefer, F.; Kliem, S. [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstrasse 400, D-01328 Dresden (Germany)

    2012-07-01

    The reactor safety issues are of primary importance for preserving the health of the population and ensuring no release of radioactivity and fission products into the environment. A part of the nuclear research focuses on improvement of the safety of existing nuclear power plants. Studies, research and efforts are a continuing process at improving the safety and reliability of existing and newly developed nuclear power plants at prevention of a core melt accident. Station blackout (loss of AC power supply) is one of the dominant accidents taken into consideration at performing accident analysis. In case of multiple failures of safety systems it leads to a severe accident. To prevent an accident to turn into a severe one or to mitigate the consequences, accident management measures must be performed. The present paper outlines possibilities for application and optimization of accident management measures following a station blackout accident. Assessed is the behaviour of the nuclear power plant during a station blackout accident without accident management measures and with application of primary/secondary side oriented accident management measures. Discussed are the possibilities for operators ' intervention and the influence of the performed accident management measures on the course of the accident. Special attention has been paid to the effectiveness of the passive feeding and physical phenomena having an influence on the system behaviour. The performed simulations show that the effectiveness of the secondary side feeding procedure can be limited due to an early evaporation or flashing effects in the feed water system. The analyzed cases show that the effectiveness of the accident management measures strongly depends on the initiation criteria applied for depressurization of the reactor coolant system. (authors)

  8. Proceedings of the US Nuclear Regulatory Commission fifteenth water reactor safety information meeting: Volume 6, Decontamination and decommissioning, accident management, TMI-2

    SciTech Connect (OSTI)

    Weiss, A. J.

    1988-02-01

    This six-volume report contains 140 papers out of the 164 that were presented at the Fifteenth Water Reactor Safety Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 26-29, 1987. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. This report, Volume 6, discusses decontamination and decommissioning, accident management, and the Three Mile Island-2 reactor accident. Thirteen reports have been cataloged separately.

  9. Accident source terms for pressurized water reactors with high-burnup cores calculated using MELCOR 1.8.5.

    SciTech Connect (OSTI)

    Gauntt, Randall O.; Powers, Dana Auburn; Ashbaugh, Scott G.; Leonard, Mark Thomas; Longmire, Pamela

    2010-04-01

    In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs2MoO4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU analyses. Additionally, current analyses suggest that the NUREG-1465 release fractions are conservative by about a factor of 2 in terms of release fractions and that release durations for in-vessel and late in-vessel release periods are in fact longer than the NUREG-1465 durations. It is currently planned that a subsequent report will further characterize these results using more refined statistical methods, permitting a more precise reformulation of the NUREG-1465 alternative source term for both LBU and HBU fuels, with the most important finding being that the NUREG-1465 formula appears to embody significant conservatism compared to current best-estimate analyses.

  10. Methanation assembly using multiple reactors

    DOE Patents [OSTI]

    Jahnke, Fred C.; Parab, Sanjay C.

    2007-07-24

    A methanation assembly for use with a water supply and a gas supply containing gas to be methanated in which a reactor assembly has a plurality of methanation reactors each for methanating gas input to the assembly and a gas delivery and cooling assembly adapted to deliver gas from the gas supply to each of said methanation reactors and to combine water from the water supply with the output of each methanation reactor being conveyed to a next methanation reactor and carry the mixture to such next methanation reactor.

  11. Catalytic reactor

    DOE Patents [OSTI]

    Aaron, Timothy Mark (East Amherst, NY); Shah, Minish Mahendra (East Amherst, NY); Jibb, Richard John (Amherst, NY)

    2009-03-10

    A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

  12. Oxidation of Zircaloy Fuel Cladding in Water-Cooled Nuclear Reactors

    SciTech Connect (OSTI)

    Digby Macdonald; Mirna Urquidi-Macdonald; Yingzi Chen; Jiahe Ai; Pilyeon Park; Han-Sang Kim

    2006-12-12

    Our work involved the continued development of the theory of passivity and passivity breakdown, in the form of the Point Defect Model, with emphasis on zirconium and zirconium alloys in reactor coolant environments, the measurement of critically-important parameters, and the development of a code that can be used by reactor operators to actively manage the accumulation of corrosion damage to the fuel cladding and other components in the heat transport circuits in both BWRs and PWRs. In addition, the modified boiling crevice model has been further developed to describe the accumulation of solutes in porous deposits (CRUD) on fuel under boiling (BWRs) and nucleate boiling (PWRs) conditions, in order to accurately describe the environment that is contact with the Zircaloy cladding. In the current report, we have derived expressions for the total steady-state current density and the partial anodic and cathodic current densities to establish a deterministic basis for describing Zircaloy oxidation. The models are “deterministic” because the relevant natural laws are satisfied explicitly, most importantly the conversation of mass and charge and the equivalence of mass and charge (Faraday’s law). Cathodic reactions (oxygen reduction and hydrogen evolution) are also included in the models, because there is evidence that they control the rate of the overall passive film formation process. Under open circuit conditions, the cathodic reactions, which must occur at the same rate as the zirconium oxidation reaction, are instrumental in determining the corrosion potential and hence the thickness of the barrier and outer layers of the passive film. Controlled hydrodynamic methods have been used to measure important parameters in the modified Point Defect Model (PDM), which is now being used to describe the growth and breakdown of the passive film on zirconium and on Zircaloy fuel sheathing in BWRs and PWRs coolant environments. The modified PDMs recognize the existence of a thick oxide outer layer over a thin barrier layer. From thermodynamic analysis, it is postulated that a hydride barrier layer forms under PWR coolant conditions whereas an oxide barrier layer forms under BWR primary coolant conditions. Thus, the introduction of hydrogen into the solution lowers the corrosion potential of zirconium to the extent that the formation of ZrH2 is predicted to be spontaneous rather than the ZrO2. Mott-Schottky analysis shows that the passive film formed on zirconium is n-type, which is consistent with the PDM, corresponding to a preponderance of oxygen/hydrogen vacancies and/or zirconium interstitials in the barrier layer. The model parameter values were extracted from electrochemical impedance spectroscopic data for zirconium in high temperature, de-aerated and hydrogenated environments by optimization. The results indicate that the corrosion resistance of zirconium is dominated by the porosity and thickness of the outer layer for both cases. The impedance model based on the PDM provides a good account of the growth of the bi-layer passive films described above, and the extracted model parameter values might be used, for example, for predicting the accumulation of general corrosion damage to Zircaloy fuel sheath in BWR and PWR operating environments. Transients in current density and film thickness for passive film formation on zirconium in dearated and hydrogenated coolant conditions have confirmed that the rate law afforded by the Point Defect Model (PDM) adequately describes the growth and thinning of the passive film. The experimental results demonstrate that the kinetics of oxygen or hydrogen vacancy generation at the metal/film interface control the rate of film growth, when the potential is displaced in the positive direction, whereas the kinetics of dissolution of the barrier layer at the barrier layer/solution interface control the rate of passive film thinning when the potential is stepped in the negative direction. In addition, the effects of second phase particles (SPPs) on the electrochemistry of passive zirconium in the

  13. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, Progress Report for Work Through September 2002, 4th Quarterly Report

    SciTech Connect (OSTI)

    Mac Donald, Philip Elsworth

    2002-09-01

    The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If no additional moderator is added to the fuel rod lattice, it is possible to attain fast neutron energy spectrum conditions in a supercritical water-cooled reactor (SCWR). This type of core can make use of either fertile or fertile-free fuel and retain a hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity. One can also add moderation and design a thermal spectrum SCWR. The Generation IV Roadmap effort has identified the thermal spectrum SCWR (followed by the fast spectrum SCWR) as one of the advanced concepts that should be developed for future use. Therefore, the work in this NERI project is addressing both types of SCWRs.

  14. Propellant actuated nuclear reactor steam depressurization valve

    DOE Patents [OSTI]

    Ehrke, Alan C. (San Jose, CA); Knepp, John B. (San Jose, CA); Skoda, George I. (Santa Clara, CA)

    1992-01-01

    A nuclear fission reactor combined with a propellant actuated depressurization and/or water injection valve is disclosed. The depressurization valve releases pressure from a water cooled, steam producing nuclear reactor when required to insure the safety of the reactor. Depressurization of the reactor pressure vessel enables gravity feeding of supplementary coolant water through the water injection valve to the reactor pressure vessel to prevent damage to the fuel core.

  15. Methods to Study Intervention Sustainability Using Pre-existing, Community Interventions: Examples from the Water, Sanitation and Hygiene Sector

    E-Print Network [OSTI]

    Arnold, Benjamin Ford

    2009-01-01

    rainwater harvesting or purchasing water from a tanker orsafe water storage, sanitation, and rainwater harvesting.safe water storage, sanitation, and rainwater harvesting.

  16. B Reactor Tour Registration Opens March 2 - Visitors Have Come...

    Energy Savers [EERE]

    and 21. Visitors will see the front face of the reactor, fan ventilation rooms, water valve pit, water process laboratories, accumulator room, and the reactor's control room. In...

  17. A Parametric Study of the Thermal-Hydraulic Response of Supercritical Light Water Reactors During Loss-of-Feedwater and Turbine-Trip Events

    SciTech Connect (OSTI)

    Cliff B. Davis; Jacopo Buongiorno; Philip E. MacDonald

    2003-09-01

    The Idaho National Engineering and Environmental Laboratory in investigating the feasibility of supercritical light water reactors for low-cost electric power production through a Nuclear Energy Research Initiative Project sponsored by the United State Department of Energy. The project is evaluating a variety of technical issues related to the fuel and reactor design, material corrosion, and safety characteristics. This paper presents the results of parametric calculations using the RELAP5 computer code to characterize the thermal-hydraulic response of supercritical reactors to transients initiated by loss-of-feedwater and turbine-trip events. The purpose of the calculations was to aid in the design of the safety systems by determining the time available for the safety systems to respond and their required capacities.

  18. Robust Low-Cost Water-Gas Shift Membrane Reactor for High-Purity Hydrogen Production form Coal-Derived Syngas

    SciTech Connect (OSTI)

    James Torkelson; Neng Ye; Zhijiang Li; Decio Coutinho; Mark Fokema

    2008-05-31

    This report details work performed in an effort to develop a low-cost, robust water gas shift membrane reactor to convert coal-derived syngas into high purity hydrogen. A sulfur- and halide-tolerant water gas shift catalyst and a sulfur-tolerant dense metallic hydrogen-permeable membrane were developed. The materials were integrated into a water gas shift membrane reactor in order to demonstrate the production of >99.97% pure hydrogen from a simulated coal-derived syngas stream containing 2000 ppm hydrogen sulfide. The objectives of the program were to (1) develop a contaminant-tolerant water gas shift catalyst that is able to achieve equilibrium carbon monoxide conversion at high space velocity and low steam to carbon monoxide ratio, (2) develop a contaminant-tolerant hydrogen-permeable membrane with a higher permeability than palladium, (3) demonstrate 1 L/h purified hydrogen production from coal-derived syngas in an integrated catalytic membrane reactor, and (4) conduct a cost analysis of the developed technology.

  19. Sustainability at BPA 2012

    SciTech Connect (OSTI)

    2012-12-01

    BPA’s Sustainability Action Plan is grounded in our commitment to environmental stewardship and Executive Order 13514 that calls on the federal agencies to “lead by example” by setting a 2020 greenhouse gas emissions target, increasing energy efficiency; reducing fleet petroleum consumption; conserving water; reducing waste; supporting sustainable communities; and leveraging federal purchasing power to promoting environmentally responsible products and technologies.

  20. Sustainability Statement Environment Team

    E-Print Network [OSTI]

    Metropolitan University www.mmu.ac.uk/environment MMU Annual Sustainability Statement 2008/2009 CO2 2 Contents 6 MMU Environment Strategy 7 Key Area 1 ­ Environmental Management Systems & Reporting 8 ­ 9 Key Area 2 ­ Energy and Carbon Emissions 10 ­ 11 Key Area 3 ­ Sustainable Buildings 12 Key Area 4 ­ Water

  1. Analysis of palladium coatings to remove hydrogen isotopes from zirconium fuel rods in Canada deuterium uranium-pressurized heavy water reactors; Thermal and neutron diffusion effects

    SciTech Connect (OSTI)

    Stokes, C.L.; Buxbaum, R.E. )

    1992-05-01

    This paper reports that, in pressurized heavy water nuclear reactors of the type standardly used in Canada (Canada deuterium uranium-pressurized heavy water reactors), the zirconium alloy pressure tubes of the core absorb deuterium produced by corrosion reactions. This deuterium weakens the tubes through hydrogen embrittlement. Thin palladium coatings on the outside of the zirconium are analyzed as a method for deuterium removal. This coating is expected to catalyze the reaction D{sub 2} + 1/2O{sub 2} {r reversible} D{sub 2}O when O{sub 2} is added to the annular (insulating) gas in the tubes. Major reductions in the deuterium concentration and, hence, hydrogen embrittlement are predicted. Potential problems such as plating the tube geometry, neutron absorption, catalyst deactivation, radioactive waste production, and oxygen corrosion are shown to be manageable. Also, a simple set of equations are derived to calculate the effect on diffusion caused by neutron interactions. Based on calculations of ordinary and neutron flux induced diffusion, a palladium coating of 1 {times} 10{sup {minus}6} m is recommended. This would cost approximately $60,000 per reactor unit and should more than double reactor lifetime. Similar coatings and similar interdiffusion calculations might have broad applications.

  2. Preliminary analysis of the postulated changes needed to achieve rail cask handling capabilities at selected light water reactors

    SciTech Connect (OSTI)

    Konzek, G.J.

    1986-02-01

    Reactor-specific railroad and crane information for all LWRs in the US was extracted from current sources of information. Based on this information, reactors were separated into two basic groups consisting of reactors with existing, usable rail cask capabilities and those without these capabilities. The latter group is the main focus of this study. The group of reactors without present rail cask handling capabilities was further separated into two subgroups consisting of reactors considered essentially incapable of handling a large rail cask of about 100 tons and reactors where postulated facility changes could result in rail cask handling capabilities. Based on a selected population of 127 reactors, the results of this assessment indicate that usable rail cask capabilities exist at 83 (65%) of the reactors. Twelve (27%) of the remaining 44 reactors are deemed incapable of handling a large rail cask without major changes, and 32 reactors are considered likely candidates for potentially achieving rail cask handling capabilities. In the latter group, facility changes were postulated that would conceptually enable these reactors to handle large rail casks. The estimated cost per plant of required facility changes varied widely from a high of about $35 million to a low of <$0.3 million. Only 11 of the 32 plants would require crane upgrades. Spur track and right-of-way costs would apparently vary widely among sites. These results are based on preliminary analyses using available generic cost data. They represent lower bound values that are useful for developing an initial assessment of the viability of the postulated changes on a system-wide basis, but are not intended to be absolute values for specific reactors or sites.

  3. Simulation of in-core neutron noise measurements for axial void profile reconstruction in boiling water reactors

    SciTech Connect (OSTI)

    Dykin, V.; Pazsit, I.

    2012-07-01

    A possibility to reconstruct the axial void profile from the simulated in-core neutron noise which is caused by density fluctuations in a Boiling Water Reactor (BWR) heated channel is considered. For this purpose, a self-contained model of the two-phase flow regime is constructed which has quantitatively and qualitatively similar properties to those observed in real BWRs. The model is subsequently used to simulate the signals of neutron detectors induced by the corresponding perturbations in the flow density. The bubbles are generated randomly in both space and time using Monte-Carlo techniques. The axial distribution of the bubble production is chosen such that the mean axial void fraction and void velocity follow the actual values of BWRs. The induced neutron noise signals are calculated and then processed by the standard signal analysis methods such as Auto-Power Spectral Density (APSD) and Cross-Power Spectral Density (CPSD). Two methods for axial void and velocity profiles reconstruction are discussed: the first one is based on the change of the break frequency of the neutron auto-power spectrum with axial core elevation, while the second refers to the estimation of transit times of propagating steam fluctuations between different axial detector positions. This paper summarizes the principles of the model and presents a numerical testing of the qualitative applicability to estimate the required parameters for the reconstruction of the void fraction profile from the neutron noise measurements. (authors)

  4. Experimental Investigation of the Root Cause Mechanism and Effectiveness of Mitigating Actions for Axial Offset Anomaly in Pressurized Water Reactors

    SciTech Connect (OSTI)

    Said Abdel-Khalik

    2005-07-02

    Axial offset anomaly (AOA) in pressurized water reactors refers to the presence of a significantly larger measured negative axial offset deviation than predicted by core design calculations. The neutron flux depression in the upper half of high-power rods experiencing significant subcooled boiling is believed to be caused by the concentration of boron species within the crud layer formed on the cladding surface. Recent investigations of the root-cause mechanism for AOA [1,2] suggest that boron build-up on the fuel is caused by precipitation of lithium metaborate (LiBO2) within the crud in regions of subcooled boiling. Indirect evidence in support of this hypothesis was inferred from operating experience at Callaway, where lithium return and hide-out were, respectively, observed following power reductions and power increases when AOA was present. However, direct evidence of lithium metaborate precipitation within the crud has, heretofore, not been shown because of its retrograde solubility. To this end, this investigation has been undertaken in order to directly verify or refute the proposed root-cause mechanism of AOA, and examine the effectiveness of possible mitigating actions to limit its impact in high power PWR cores.

  5. Materials Reliability Program Resistance to Primary Water Stress Corrosion Cracking of Alloys 690, 52, and 152 in Pressurized Water Reactors (MRP-111)

    SciTech Connect (OSTI)

    H.Xu, S.Fyfitch, P.Scott, M.Foucault, R.Kilian, and M.Winters

    2004-03-01

    Over the last thirty years, stress corrosion cracking in PWR primary water (PWSCC) has been observed in numerous Alloy 600 component items and associated welds, sometimes after relatively long incubation times. Repairs and replacements have generally utilized wrought Alloy 690 material and its compatible weld metals (Alloy 152 and Alloy 52), which have been shown to be very highly resistant to PWSCC in laboratory experiments and have been free from cracking in operating reactors over periods already up to nearly 15 years. It is nevertheless prudent for the PWR industry to attempt to quantify the longevity of these materials with respect to aging degradation by corrosion in order to provide a sound technical basis for the development of future inspection requirements for repaired or replaced component items. This document first reviews numerous laboratory tests, conducted over the last two decades, that were performed with wrought Alloy 690 and Alloy 52 or Alloy 152 weld materials under various test conditions pertinent to corrosion resistance in PWR environments. The main focus of the present review is on PWSCC, but secondary-side conditions are also briefly considered.

  6. An Environmental and Economic Trade-off Analysis of Manufacturing Process Chains to Inform Decision Making for Sustainability

    E-Print Network [OSTI]

    Robinson, Stefanie L.

    2013-01-01

    quantifies the sustainability impacts (energy, water, andall of the sustainability impacts (energy, water, and waste)quantifying the sustainability impacts ( energy consumption,

  7. Sustainability Around the House | Department of Energy

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

    Sustainability Around the House Sustainability Around the House April 13, 2015 - 7:46am Addthis Rain barrels collect rain water and provide a free source of fresh water for your...

  8. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 1. Main report. Technical report, September 1977-October 1979

    SciTech Connect (OSTI)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE.

  9. Mechanism of Irradiation Assisted Cracking of Core Components in Light Water Reactors

    SciTech Connect (OSTI)

    Gary S. Was; Michael Atzmon; Lumin Wang

    2003-04-28

    The overall goal of the project is to determine the mechanism of irradiation assisted stress corrosion cracking (IASCC). IASCC has been linked to hardening, microstructural and microchemical changes during irradiation. Unfortunately, all of these changes occur simultaneously and at similar rates during irradiation, making attribution of IASCC to any one of these features nearly impossible to determine. The strategy set forth in this project is to develop means to separate microstructural from microchemical changes to evaluate each separately for their effect on IASCC. In the first part, post irradiation annealing (PIA) treatments are used to anneal the irradiated microstructure, leaving only radiation induced segregation (RIS) for evaluation for its contribution to IASCC. The second part of the strategy is to use low temperature irradiation to produce a radiation damage dislocation loop microstructure without radiation induced segregation in order to evaluate the effect of the dislocation microstructure alone. A radiation annealing model was developed based on the elimination of dislocation loops by vacancy absorption. Results showed that there were indeed, time-temperature annealing combinations that leave the radiation induced segregation profile largely unaltered while the dislocation microstructure is significantly reduced. Proton irradiation of 304 stainless steel irradiated with 3.2 MeV protons to 1.0 or 2.5 dpa resulted in grain boundary depletion of chromium and enrichment of nickel and a radiation damaged microstructure. Post irradiation annealing at temperatures of 500 ? 600°C for times of up to 45 min. removed the dislocation microstructure to a greater degree with increasing temperatures, or times at temperature, while leaving the radiation induced segregation profile relatively unaltered. Constant extension rate tensile (CERT) experiments in 288°C water containing 2 ppm O2 and with a conductivity of 0.2 mS/cm and at a strain rate of 3 x 10-7 s-1 showed that the IASCC susceptibility, as measured by the crack length per unit strain, decreased with very short anneals and was almost completely removed by an anneal at 500°C for 45 min. This annealing treatment removed about 15% of the dislocation microstructure and the irradiation hardening, but did not affect the grain boundary chromium depletion or nickel segregation, nor did it affect the grain boundary content of other minor impurities. These results indicate that RIS is not the sole controlling feature of IASCC in irradiated stainless steels in normal water chemistry. The isolation of the irradiated microstructure was approached using low temperature irradiation or combinations of low and high temperature irradiations to achieve a stable, irradiated microstructure without RIS. Experiments were successful in achieving a high degree of irradiation hardening without any evidence of RIS of either major or minor elements. The low temperature irradiations to doses up to 0.3 dpa at T<75°C were also very successful in producing hardening to levels considerably above that for irradiations conducted under nominal conditions of 1 dpa at 360°C. However, the microstructure consisted of an extremely fine dispersion of defect clusters of sizes that are not resolvable by either transmission electron microscopy (TEM) or small angle x-ray scattering (SAXS). The microstructure was not stable at the 288°C IASCC test temperature and resulted in rapid reduction of hardening and presumably, annealing of the defect clusters at this temperature as well. Nevertheless, the annealing studies showed that treatments that resulted in significant decreases in the hardening produced small changes in the dislocation microstructure that were confined to the elimination of the finest of loops (~1 nm). These results substantiate the importance of the very fine defect microstructure in the IASCC process. The results of this program provide the first definitive evidence that RIS is not the sole controlling factor in the irradiation assisted stress corrosion cracking of austenitic stain

  10. Sustainable water resources development in Kuwait : an integrated approach with comparative analysis of the case of Singapore

    E-Print Network [OSTI]

    Nazerali, Nasruddin A

    2007-01-01

    This thesis assesses the water resource status of Kuwait and Singapore, both countries considered as water scarce. The institutional aspect of Integrated Water Resource Management (IWRM) efforts in both countries is closely ...

  11. Co-Designing Sustainable Communities: The Identification and Incorporation of Social Performance Metrics in Native American Sustainable Housing and Renewable Energy System Design

    E-Print Network [OSTI]

    Shelby, Ryan

    2013-01-01

    of Renewable Energy and Sustainability (CARES) in the hopeof Renewable Energy and Sustainability (CARES) project (withor save energy and water” and address the “sustainability….

  12. Sustainable Development

    E-Print Network [OSTI]

    Brierley, Andrew

    ) Sustainable Development Contributing Schools Biology; Chemistry; Computer Science; Geography & Geosciences enquiries Dr Emilia Ferraro, Department of Geography & Sustainable Development E: gg. Resources Very few universities currently offer undergraduate degrees in Sustainable Development. Uniquely

  13. Departmental Sustainability

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2015-12-17

    The order defines requirements and responsibilities for managing sustainability DOE to ensure that the Department carries out its missions in a sustainable manner that addresses national energy security and global environmental challenges, and advances sustainable, efficient and reliable energy for the future; institute wholesale cultural change to factor sustainability and greenhouse gas (GHG) reductions into all DOE corporate management decisions; and ensure that DOE achieves the sustainability goals established in its Strategic Sustainability Performance Plan.

  14. Sustainable Transport

    E-Print Network [OSTI]

    Webber, Melvin

    2006-01-01

    THOUGHT PIECE Sustainable Transport by Melvin M. Webberwant to sustain any mode of transport only if we judge it todraconian in rejecting transport modes that have failed in

  15. ANNUAL SUSTAINABILITY REPORT YORK UNIVERSITY

    E-Print Network [OSTI]

    and opportunities associated with sustainability and to propose innovative ideas and solutions that will help us staff members working on sustainability have helped us to achieve truly transformational results in energy reduction, waste diversion, water conservation, local food and sustainable transportation. In fact

  16. TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

    SciTech Connect (OSTI)

    Liles, D.R.; Mahaffy, J.H.

    1986-07-01

    The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients.

  17. TITAN : an advanced three dimensional coupled neutronicthermal-hydraulics code for light water nuclear reactor core analysis

    E-Print Network [OSTI]

    Griggs, D. P.

    1984-01-01

    The accurate analysis of nuclear reactor transients frequently requires that neutronics, thermal-hydraulics and feedback be included. A number of coupled neutronics/thermal-hydraulics codes have been developed for this ...

  18. Small modular reactors (SMRs) such...

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

    (ARO), using soluble boron in the coolant for reactivity control. Conversely, boiling water reactors (BWRs) typically maneuver their control blades as often as every 2 GWdmtU...

  19. Post-remedial-action survey report for Kinetic Experiment Water Boiler Reactor Facility, Santa Susana Field Laboratories, Rockwell International, Ventura County, California

    SciTech Connect (OSTI)

    Wynveen, R.A.; Smith, W.H.; Sholeen, C.M.; Flynn, K.F.; Justus, A.L.

    1981-10-01

    Rockwell International's Santa Susana Laboratories in Ventura County, California, have been the site of numerous federally-funded contracted projects involving the use of radioactive materials. Among these was the Kinetics Experiment Water Boiler (KEWB) Reactor which was operated under the auspices of the US Atomic Energy Commission (AEC). The KEWB Reactor was last operated in 1966. The facility was subsequently declared excess and decontamination and decommissioning operations were conducted during the first half of calendar year 1975. The facility was completely dismantled and the site graded to blend with the surrounding terrain. During October 1981, a post-remedial-action (certification) survey of the KEWB site was conducted on the behalf of the US Department of Energy by the Radiological Survey Group (RSG) of the Occupational Health and Safety Division's Health Physics Section (OHS/HP) of Argonne National Laboratory (ANL). The survey confirmed that the site was free from contamination and could be released for unrestricted use.

  20. Boiling Water Reactor Fuel Behavior Under Reactivity-Initiated-Accident Conditions at Burnup of 41 to 45 GWd/tonne U

    SciTech Connect (OSTI)

    Nakamura, Takehiko; Yoshinaga, Makio; Takahashi, Masato; Okonogi, Kazunari; Ishijima, Kiyomi

    2000-02-15

    Boiling water reactor (BWR) fuel at burnup of 41 to 45 GWd/tonne U was pulse irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity-initiated-accident conditions. Current Japanese BWR fuel, 8 x 8BJ type (Step I), from Fukushima-Daiichi Unit 3 was refabricated into short segments, and the test rods were promptly subjected to thermal energy from 293 to 607 J/g (70 to 145 cal/g) within {approx}20 ms. The fuel cladding was ductile enough to survive the prompt deformation due to pellet cladding mechanical interaction, while the plastic hoop strain reached 1.5% at the peak location. Transient fission gas release by the pulse irradiation varied from 3.1 to 8.2%, depending on the peak fuel enthalpy and the steady-state operation conditions.

  1. Transactions of the twenty-third water reactor safety information meeting to be held at Bethesda Marriott Hotel, Bethesda, Maryland, October 23--25, 1995

    SciTech Connect (OSTI)

    Monteleone, S. [comp.

    1995-09-01

    This report contains summaries of papers on reactor safety research to be presented at the 23rd Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23--25, 1995. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory, Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting and are given in the order of their presentation in each session.

  2. Monitoring and evaluation of household water treatment and safe storage technologies : the sustained use of the KOSIM ceramic water filter in northern region Ghana

    E-Print Network [OSTI]

    Clopeck, Katherine L. (Katherine Lorraine)

    2009-01-01

    Today, approximately 884 million people lack access to an improved drinking water (WHO and UNICEF, 2008). According to the World Health Organization (WHO), contaminated water and poor sanitation cause 30,000 deaths worldwide ...

  3. SustainabilityStudies sustainability@uci.edu

    E-Print Network [OSTI]

    Rose, Michael R.

    SustainabilityStudies sustainability@uci.edu Sustainable Business Management Certificate Program in the sustainability field, such as managers, directors, architects, urban planners, landscape architects, designers opportunities and risks while minimizing potential liability. extension.uci.edu/sustainability Sustainable

  4. USING LIGA BASED MICROFABRICATION TO IMPROVE OVERALL HEAT TRANSFER EFFICIENCY OF PRESSURIZED WATER REACTOR: I. Effects of Different Micro Pattern on Overall Heat Transfer.

    SciTech Connect (OSTI)

    Zhang, M.; Ibekwe, S.; Li, G.; Pang, S.S.; and Lian, K.

    2006-07-01

    The Pressurized Water Reactors (PWRs in Figure 1) were originally developed for naval propulsion purposes, and then adapted to land-based applications. It has three parts: the reactor coolant system, the steam generator and the condenser. The Steam generator (a yellow area in Figure 1) is a shell and tube heat exchanger with high-pressure primary water passing through the tube side and lower pressure secondary feed water as well as steam passing through the shell side. Therefore, a key issue in increasing the efficiency of heat exchanger is to improve the design of steam generator, which is directly translated into economic benefits. The past research works show that the presence of a pin-fin array in a channel enhances the heat transfer significantly. Hence, using microfabrication techniques, such as LIGA, micro-molding or electroplating, some special microstructures can be fabricated around the tubes in the heat exchanger to increase the heat-exchanging efficiency and reduce the overall size of the heat-exchanger for the given heat transfer rates. In this paper, micro-pin fins of different densities made of SU-8 photoresist are fabricated and studied to evaluate overall heat transfer efficiency. The results show that there is an optimized micro pin-fin configuration that has the best overall heat transfer effects.

  5. Departmental Sustainability

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2011-05-02

    The order defines requirements and responsibilities for managing sustainability DOE to ensure that the Department carries out its missions in a sustainable manner that addresses national energy security and global environmental challenges, and advances sustainable, efficient and reliable energy for the future; institute wholesale cultural change to factor sustainability and greenhouse gas (GHG) reductions into all DOE corporate management decisions; and ensure that DOE achieves the sustainability goals established in its Strategic Sustainability Performance Plan. Supersedes DOE O 450.1A and DOE O 430.2B

  6. Nuclear reactor reflector

    DOE Patents [OSTI]

    Hopkins, Ronald J. (Pensacola, FL); Land, John T. (Pensacola, FL); Misvel, Michael C. (Pensacola, FL)

    1994-01-01

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.

  7. Nuclear reactor reflector

    DOE Patents [OSTI]

    Hopkins, R.J.; Land, J.T.; Misvel, M.C.

    1994-06-07

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled. 12 figs.

  8. Neutronic evaluation of a non-fertile fuel for the disposition of weapons-grade plutonium in a boiling water reactor

    SciTech Connect (OSTI)

    Sterbentz, J.W.

    1994-10-01

    A new non-fertile, weapons-grade plutonium oxide fuel concept is developed and evaluated for deep burn applications in a boiling water reactor environment using the General Electric 8x8 Advanced Boiling Water Reactor (ABWR) fuel assembly dimensions and pitch. Detailed infinite lattice fuel burnup results and neutronic performance characteristics are given and although preliminary in nature, clearly demonstrate the fuel`s potential as an effective means to expedite the disposition of plutonium in existing light water reactors. The new non-fertile fuel concept is an all oxide composition containing plutonia, zirconia, calcia, and erbia having the following design weight percentages: 8.3; 80.4; 9.7; and 1.6. This fuel composition in an infinite fuel lattice operating at linear heat generation rates of 6.0 or 12.0 kW/ft per rod can remain critical for up to 1,200 and 600 Effective Full Power Days (EFPD), respectively, and achieve a burnup of 7.45 {times} 10{sup 20} f/cc. These burnups correspond to a 71--73% total plutonium isotope destruction and a 91--94% destruction of the {sup 239}Pu isotope for the 0--40% moderator steam void condition. Total plutonium destruction greater than 73% is possible with a fuel management scheme that allows subcritical fuel assemblies to be driven by adjacent high reactivity assemblies. The fuel exhibits very favorable neutron characteristics from beginning-of-life (BOL) to end-of-life (EOL). Prompt fuel Doppler coefficient of reactivity are negative, with values ranging between {minus}0.4 to {minus}2.0 pcm/K over the temperature range of 900 to 2,200 K. The ABWR fuel lattice remains in an undermoderated condition for both hot operational and cold startup conditions over the entire fuel burnup lifetime.

  9. Methods to Study Intervention Sustainability Using Pre-existing, Community Interventions: Examples from the Water, Sanitation and Hygiene Sector

    E-Print Network [OSTI]

    Arnold, Benjamin Ford

    2009-01-01

    sanitation, and rainwater harvesting. Epidemiol Infect.sanitation, and rainwater harvesting. Epidemiol Infect.and none report rainwater harvesting or purchasing water

  10. In-situ Surface Enhanced Raman Spectroscopy Investigation of the Surface Films on Alloy 600 and Alloy 690 in Pressurized Water Reactor-Primary Water

    E-Print Network [OSTI]

    Wang, Feng

    2012-01-01

    oxidation of Alloy 600 in PWR Primary Water. The layered-oxidation of Alloy 690 in PWR Primary Water. The film ofwith oxidation of Alloy 600 in PWR Primary Water. The film

  11. Twenty-second water reactor safety information meeting. Volume 2: Severe accident research, thermal hydraulic research for advanced passive LWRs, high-burnup fuel behavior

    SciTech Connect (OSTI)

    Monteleone, S. [comp.

    1995-04-01

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24-26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting.

  12. A Blueprint for Urban Sustainability: Integrating Sustainable...

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

    A Blueprint for Urban Sustainability: Integrating Sustainable Energy Practices into Metropolitan Planning, May 2004 A Blueprint for Urban Sustainability: Integrating Sustainable...

  13. SCDAP/RELAP5/MOD 3.1 code manual: MATPRO, A library of materials properties for Light-Water-Reactor accident analysis. Volume 4

    SciTech Connect (OSTI)

    Hagrman, D.T.; Allison, C.M.; Berna, G.A.

    1995-06-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light -- water-reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume, Volume IV, describes the material properties correlations and computer subroutines (MATPRO) used by SCDAP/RELAP5. formulation of the materials properties are generally semi-empirical in nature. The materials property subroutines contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, cadmium, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, fill gas mixtures, carbon steel, and tungsten. This document also contains descriptions of the reaction and solution rate models needed to analyze a reactor accident.

  14. sustainable environment

    E-Print Network [OSTI]

    change · Build management skills for policy development sustainable resource management Cefas helpssustainable resource management environment fisheries aquaculture Cefas capability statement #12;sustainable resource management 4 managing environmental resources 4 institutional development 5 an integrated

  15. VirginiaTechCenterfor SustainableNanotechnolgy

    E-Print Network [OSTI]

    Virginia Tech

    wastewater treatment plant. These results confirm for the first time that sludge disposal or the use Sustainable Energy Renewable Materials Sustainable Water Cognition and Communication National Security, sustainable energy, green manufacturing, and the production of clean drinking water, just to name a few

  16. Westinghouse-CASL team wins major computing award for reactor...

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

    performed core physics simulations of the Westinghouse AP1000 pressurized water reactor (PWR) core using CASL's Virtual Environment for Reactor Application (VERA). Westinghouse is...

  17. ASSESSMENT OF SMALL AND MODULAR REACTOR NUCLEAR FUEL COST 

    E-Print Network [OSTI]

    Pannier, Christopher 1992-

    2012-05-03

    INCAS INtegrated model for the Competitiveness Analysis of Small modular reactors LWR Light Water Reactor NEI Nuclear Energy Institute PWR Pressurized Water Reactor PHWR Pressurized Heavy Water Reactor SEMER Système d’Évaluation et de Modélisation... ...................................................... 27 8 LWR Fuel Cost ..................................................................................................... 28 9 SMR Fuel Cost ..................................................................................................... 29...

  18. Nuclear Power Trends Energy Economics and Sustainability

    E-Print Network [OSTI]

    Nuclear Power Trends Energy Economics and Sustainability L. H. Tsoukalas Purdue University Nuclear Nuclear Today · 439 nuclear power reactors (31 countries) · Over 12,000 years of operating experience · Nuclear reactors supply 16% of the world's electricity as base-load power (372,000 MWe of total capacity

  19. sustainable school of sustainable

    E-Print Network [OSTI]

    systems engineering School Director Ronald G. Askin Personalized learning Educational gaming Energy and product lifecycle analysis Earth systems engineering Water purification Resource-climate interactions Underground Infrastructure Construction Management school of electrical, computer and energy engineering

  20. A study of the factors affecting the sustainability of community managed rural water supply schemes in Sri Lanka

    E-Print Network [OSTI]

    Amerasinghe, Nishanthi Manjula

    2009-01-01

    Disparities in water supply coverage in urban and rural areas are high in developing countries, with rural coverage being much lower. The inability of governments to provide the service because of resource constraints, and ...

  1. Preliminary results of calculations for heavy-water nuclear-power-plant reactors employing {sup 235}U, {sup 233}U, and {sup 232}Th as a fuel and meeting requirements of a nonproliferation of nuclear weapons

    SciTech Connect (OSTI)

    Ioffe, B. L.; Kochurov, B. P. [Institute of Theoretical and Experimental Physics (Russian Federation)

    2012-02-15

    A physical design is developed for a gas-cooled heavy-water nuclear reactor intended for a project of a nuclear power plant. As a fuel, the reactor would employ thorium with a small admixture of enriched uranium that contains not more than 20% of {sup 235}U. It operates in the open-cycle mode involving {sup 233}U production from thorium and its subsequent burnup. The reactor meets the conditions of a nonproliferation of nuclear weapons: the content of fissionable isotopes in uranium at all stages of the process, including the final one, is below the threshold for constructing an atomic bomb, the amount of product plutonium being extremely small.

  2. NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapters 2-13, project number 669

    SciTech Connect (OSTI)

    Not Available

    1994-08-01

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume I, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff`s review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

  3. A SCOPING STUDY: Development of Probabilistic Risk Assessment Models for Reactivity Insertion Accidents During Shutdown In U.S. Commercial Light Water Reactors

    SciTech Connect (OSTI)

    S. Khericha

    2011-06-01

    This report documents the scoping study of developing generic simplified fuel damage risk models for quantitative analysis from inadvertent reactivity insertion events during shutdown (SD) in light water pressurized and boiling water reactors. In the past, nuclear fuel reactivity accidents have been analyzed both mainly deterministically and probabilistically for at-power and SD operations of nuclear power plants (NPPs). Since then, many NPPs had power up-rates and longer refueling intervals, which resulted in fuel configurations that may potentially respond differently (in an undesirable way) to reactivity accidents. Also, as shown in a recent event, several inadvertent operator actions caused potential nuclear fuel reactivity insertion accident during SD operations. The set inadvertent operator actions are likely to be plant- and operation-state specific and could lead to accident sequences. This study is an outcome of the concern which arose after the inadvertent withdrawal of control rods at Dresden Unit 3 in 2008 due to operator actions in the plant inadvertently three control rods were withdrawn from the reactor without knowledge of the main control room operator. The purpose of this Standardized Plant Analysis Risk (SPAR) Model development project is to develop simplified SPAR Models that can be used by staff analysts to perform risk analyses of operating events and/or conditions occurring during SD operation. These types of accident scenarios are dominated by the operator actions, (e.g., misalignment of valves, failure to follow procedures and errors of commissions). Human error probabilities specific to this model were assessed using the methodology developed for SPAR model human error evaluations. The event trees, fault trees, basic event data and data sources for the model are provided in the report. The end state is defined as the reactor becomes critical. The scoping study includes a brief literature search/review of historical events, developments of a small set of comprehensive event trees and fault trees and recommendation for future work.

  4. NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapter 1, project number 669

    SciTech Connect (OSTI)

    Not Available

    1994-08-01

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume 1, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff`s review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

  5. The concept of the use of recycled uranium for increasing the degree of security of export deliveries of fuel for light-water reactors

    SciTech Connect (OSTI)

    Alekseev, P. N.; Ivanov, E. A.; Nevinitsa, V. A.; Ponomarev-Stepnoi, N. N.; Rumyantsev, A. N.; Shmelev, V. M. [Russian Research Center Kurchatov Institute (Russian Federation); Borisevich, V. D.; Smirnov, A. Yu.; Sulaberidze, G. A. [National Nuclear Research University MEPhI (Russian Federation)

    2010-12-15

    The present paper deals with investigation of the possibilities for reducing the risk of proliferation of fissionable materials by means of increasing the degree of protection of fresh fuel intended for light-water reactors against unsanctioned use in the case of withdrawal of a recipient country of deliveries from IAEA safeguards. It is shown that the use of recycled uranium for manufacturing export nuclear fuel makes transfer of nuclear material removed from the fuel assemblies for weapons purposes difficult because of the presence of isotope {sup 232}U, whose content increases when one attempts to enrich uranium extracted from fresh fuel. In combination with restricted access to technologies for isotope separation by means of establishing international centers for uranium enrichment, this technical measure can significantly reduce the risk of proliferation associated with export deliveries of fuel made of low-enriched uranium. The assessment of a maximum level of contamination of nuclear material being transferred by isotope {sup 232}U for the given isotope composition of the initial fuel is obtained. The concept of further investigations of the degree of security of export deliveries of fuel assemblies with recycled uranium intended for light-water reactors is suggested.

  6. 324 Building B-Cell Pressurized Water Reactor Spent Fuel Packaging & Shipment RL Readiness Assessment Final Report [SEC 1 Thru 3

    SciTech Connect (OSTI)

    HUMPHREYS, D C

    2002-08-01

    A parallel readiness assessment (RA) was conducted by independent Fluor Hanford (FH) and U. S. Department of Energy, Richland Operations Office (RL) team to verify that an adequate state of readiness had been achieved for activities associated with the packaging and shipping of pressurized water reactor fuel assemblies from B-Cell in the 324 Building to the interim storage area at the Canister Storage Building in the 200 Area. The RL review was conducted in parallel with the FH review in accordance with the Joint RL/FH Implementation Plan (Appendix B). The RL RA Team members were assigned a FH RA Team counterpart for the review. With this one-on-one approach, the RL RA Team was able to assess the FH Team's performance, competence, and adherence to the implementation plan and evaluate the level of facility readiness. The RL RA Team agrees with the FH determination that startup of the 324 Building B-Cell pressurized water reactor spent nuclear fuel packaging and shipping operations can safely proceed, pending completion of the identified pre-start items in the FH final report (see Appendix A), completion of the manageable list of open items included in the facility's declaration of readiness, and execution of the startup plan to operations.

  7. Water Resources Forests & Water

    E-Print Network [OSTI]

    Water Resources Forests & Water More than half of the nation's freshwater supply originates on forestland. Healthy and sustainable forests can help ensure a continuous supply of clean and abundant water. Not only does forestland provide the cleanest water of any land use, it also helps absorb rainfall

  8. In-situ Surface Enhanced Raman Spectroscopy Investigation of the Surface Films on Alloy 600 and Alloy 690 in Pressurized Water Reactor-Primary Water

    E-Print Network [OSTI]

    Wang, Feng

    2012-01-01

    Cook et. al. developed the Pourbaix diagram for the nickel-W.G. Cook has proposed a Pourbaix diagram for chromium-waterPWR primary water. At 300 o C, Pourbaix diagram shows a wide

  9. Metropolitan Accessibility and Transportation Sustainability:Sustainability

    E-Print Network [OSTI]

    Papalambros, Panos

    Metropolitan Accessibility and Transportation Sustainability:Sustainability: Comparative Reduce (fulfillment of)Promote Sustainability: Meet needs of (fulfillment of) needs present Institute SMART Sustainable Mobility and Accessibility Research andSMART Sustainable Mobility

  10. An Environmental and Economic Trade-off Analysis of Manufacturing Process Chains to Inform Decision Making for Sustainability

    E-Print Network [OSTI]

    Robinson, Stefanie L.

    2013-01-01

    all of the sustainability impacts (energy, water, and waste)quantifying the sustainability impacts ( energy consumption,quantifying the sustainability impacts (energy consumption,

  11. Fast Breeder Reactor studies

    SciTech Connect (OSTI)

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  12. Development of Modeling Capabilities for the Analysis of Supercritical Water-Cooled Reactor Thermal-Hydraulics and Dynamics

    SciTech Connect (OSTI)

    Dr. Michael Z. Podowski

    2009-04-16

    Develop an experimental and theoretical data base for heat transfer in tubes and channels cooled by water and CO2 at supercritical pressures.

  13. Committee for Sustainability

    E-Print Network [OSTI]

    Vetriani, Costantino

    of the nation, electricity use at Rutgers, both in total and per square foot of floor space, has increased stable since 2000. Solid waste disposal, energy usage, water usage, and emissions of greenhouse gases engagement with sustainability issues. New faculty initiatives in research and teaching on energy and climate

  14. Two-dimensional DORT discrete ordinates X-Y geometry neutron flux calculations for the Halden Heavy Boiling Water Reactor core configurations

    SciTech Connect (OSTI)

    Slater, C.O.

    1990-07-01

    Results are reported for two-dimensional discrete ordinates, X-Y geometry calculations performed for seven Halden Heavy Boiling Water Reactor core configurations. The calculations were performed in support of an effort to reassess the neutron fluence received by the reactor vessel. Nickel foil measurement data indicated considerable underprediction of fluences by the previously used multigroup removal- diffusion method. Therefore, calculations by a more accurate method were deemed appropriate. For each core configuration, data are presented for (1) integral fluxes in the core and near the vessel wall, (2) neutron spectra at selected locations, (3) isoflux contours superimposed on the geometry models, (4) plots of the geometry models, and (5) input for the calculations. The initial calculations were performed with several mesh sizes. Comparisons of the results from these calculations indicated that the uncertainty in the calculated fluxes should be less than 10%. However, three-dimensional effects (such as axial asymmetry in the fuel loading) could contribute to much greater uncertainty in the calculated neutron fluxes. 7 refs., 22 figs., 11 tabs.

  15. Venice Sustainability Advisory Panel

    E-Print Network [OSTI]

    2009-01-01

    Venice Sustainability Advisory PanelFINAL REPORT Venice Sustainability Advisory Panel FinalInvestigator The Venice Sustainability Advisory Panel (

  16. Twenty-First Water Reactor Safety Information Meeting. Volume 3, Primary system integrity; Aging research, products and applications; Structural and seismic engineering; Seismology and geology: Proceedings

    SciTech Connect (OSTI)

    Monteleone, S.

    1994-04-01

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25-27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Selected papers were indexed separately for inclusion in the Energy Science and Technology Database.

  17. Investigation of the use of nanofluids to enhance the In-Vessel Retention capabilities of Advanced Light Water Reactors

    E-Print Network [OSTI]

    Hannink, Ryan Christopher

    2007-01-01

    Nanofluids at very low concentrations experimentally exhibit a substantial increase in Critical Heat Flux (CHF) compared to water. The use of a nanofluid in the In-Vessel Retention (IVR) severe accident management strategy, ...

  18. Feasibility of Remote Nuclear Reactor Antineutrino Directionality via Elastic Electron Scattering in the WATer CHerenkov Monitor of ANtineutrinos (WATCHMAN) 

    E-Print Network [OSTI]

    Hellfeld, Daniel

    2015-05-06

    interactions, gamma rays from the PMTs, detector walls, and surrounding rock, as well as the decays of cosmogenic radionuclides and water-borne radon. Preliminary results indicate that while most of the sources of background can be adequately controlled...

  19. Syngas Production By Thermochemical Conversion Of H2o And Co2 Mixtures Using A Novel Reactor Design

    SciTech Connect (OSTI)

    Pearlman, Howard; Chen, Chien-Hua

    2014-08-27

    The Department of Energy awarded Advanced Cooling Technologies, Inc. (ACT) an SBIR Phase II contract (#DE-SC0004729) to develop a high-temperature solar thermochemical reactor for syngas production using water and/or carbon dioxide as feedstocks. The technology aims to provide a renewable and sustainable alternative to fossil fuels, promote energy independence and mitigate adverse issues associated with climate change by essentially recycling carbon from carbon dioxide emitted by the combustion of hydrocarbon fuels. To commercialize the technology and drive down the cost of solar fuels, new advances are needed in materials development and reactor design, both of which are integral elements in this program.

  20. Defective fuel rod detection in operating pressurized water reactors during periods of continuously decreasing fuel rod integrity levels

    SciTech Connect (OSTI)

    Zanker, H. )

    1989-09-01

    Periods of continuously decreasing levels of fuel rod integrity due to debris-induced cladding damage, vibration-induced fretting wear of the cladding, etc. cause difficulties in the assessment of fuel rod performance from coolant activity data. The calculational models currently in use for this purpose in nuclear power plants are not sufficiently capable of indicating cases in which they are invalid. This can mislead reactor operators by misinterpretation of the coolant activity data, especially in situations where fast reactions are necessary. A quick test of validity is suggested to check the applicability of the currently available calculational models for estimating the number and average size of fuel rod defects. This paper describes how to recognize immediately periods of continuously decreasing levels of fuel rod integrity in order to prevent complications in routine power plant maintenance as well as accident situations caused by more severe fuel rod degradation.

  1. A REPORT TO THE PRESIDENT ON Sustainability

    E-Print Network [OSTI]

    Johnston, Daniel

    A REPORT TO THE PRESIDENT ON Sustainability Presented by The University of Texas at Austin Staff to the President Prepared by The University of Texas at Austin Staff Council Annual committee on Sustainability-line sustainability initiatives such as recycling, energy conservation, water conservation, and campus landscape

  2. Site Sustainability Plan An Authentic Commitment

    E-Print Network [OSTI]

    #12;#12;Contents Site Sustainability Plan An Authentic Commitment Contents Acronyms and Abbreviations iv Executive Summary v Greenhouse Gas 1 Energy 6 High Performance Sustainable Buildings 9 Fleet 11 Water 13 Pollution Prevention 15 Sustainable Acquisition 19 Electronic Stewardship and Data Centers 20

  3. Sustainability Performance Office | Department of Energy

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

    Sustainability Performance Office Sustainability Performance Office Sustainability Performance Office Sustainability Performance Office Sustainability Performance Office...

  4. ENVIRONMENTAL SUSTAINABILITY

    E-Print Network [OSTI]

    27 ICT AND ENVIRONMENTAL SUSTAINABILITY T he environment is a large complex sys- tem. Managing. Environmental Monitoring and Associated Resource Management and Risk Mitigation ICTimprovestheabilitytoobtain,storeandinte- grate large volumes of environmental data and to conductsimulationandanalysisinrealtime

  5. Sustainable Scientists

    E-Print Network [OSTI]

    Mills, Evan

    2009-01-01

    thanks to extensive green and energy-efficient features, andenergy efficiency improvement programs (xxxviii) and their GreenGreen Grid alliance to address efficiency and sustainability issues (xxxvii). This group is exploring energy

  6. Particle Size Distributions of Debris Upstream and Downstream of the Containment Sump Strainer in a Light Water Reactor 

    E-Print Network [OSTI]

    Kappes, Matthew Jareb

    2015-07-30

    During a LOCA in a LWR, a containment-sump strainer filters debris, generated from fibrous thermal insulation, from the water collected in the containment sump. The buildup of debris on the strainer and the bypass of debris through the strainer lead...

  7. Science Serving Sustainability

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

    Buildings Greening Transportation Green Purchasing & Green Technology Pollution Prevention Science Serving Sustainability ENVIRONMENTAL SUSTAINABILITY GOALS at...

  8. High Performance Sustainable Buildings

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

    Buildings Greening Transportation Green Purchasing & Green Technology Pollution Prevention Science Serving Sustainability ENVIRONMENTAL SUSTAINABILITY GOALS at...

  9. Use of an influence diagram and fuzzy probability for evaluating accident management in a boiling water reactor

    SciTech Connect (OSTI)

    Yu, D.; Kastenberg, W.E.; Okrent, D. (Univ. of California, Los Angeles, CA (United States). Mechanical, Aerospace, and Nuclear Engineering Dept.)

    1994-06-01

    A new approach is presented for evaluating the uncertainties inherent in severe accident management strategies. At first, this analysis considers accident management as a decision problem (i.e., applying a strategy compared with do nothing) and uses an influence diagram. To evaluate imprecise node probabilities in the influence diagram, the analysis introduces the concept of a fuzzy probability. When fuzzy logic is applied, fuzzy probabilities are easily propagated to obtain results. In addition, the results obtained provide not only information similar to the classical approach, which uses point-estimate values, but also additional information regarding the impact of using imprecise input data. As an illustrative example, the proposed methodology is applied to the evaluation of the drywell flooding strategy for a long-term station blackout sequence at the Peach Bottom nuclear power plant. The results show that the drywell flooding strategy is beneficial for preventing reactor vessel breach. It is also effective for reducing the probability of containment failure for both liner melt-through and late overpressurization. Even though uncertainty exists in the results, flooding is preferred to do nothing when evaluated in terms of two risk measures: early and late fatalities.

  10. Light Water Reactor Sustainability (LWRS) Program – Non-Destructive Evaluation (NDE) R&D Roadmap for Determining Remaining Useful Life of Aging Cables in Nuclear Power Plants

    SciTech Connect (OSTI)

    Simmons, Kevin L.; Ramuhalli, Pradeep; Brenchley, David L.; Coble, Jamie B.; Hashemian, Hash; Konnik, Robert; Ray, Sheila

    2012-09-14

    The purpose of the non-destructive evaluation (NDE) R&D Roadmap for Cables is to support the Materials Aging and Degradation (MAaD) R&D pathway. The focus of the workshop was to identify the technical gaps in detecting aging cables and predicting their remaining life expectancy. The workshop was held in Knoxville, Tennessee, on July 30, 2012, at Analysis and Measurement Services Corporation (AMS) headquarters. The workshop was attended by 30 experts in materials, electrical engineering, U.S. Nuclear Regulatory Commission (NRC), U.S. Department of Energy (DOE) National Laboratories (Oak Ridge National Laboratory, Pacific Northwest National Laboratory, Argonne National Laboratory, and Idaho National Engineering Laboratory), NDE instrumentation development, universities, commercial NDE services and cable manufacturers, and Electric Power Research Institute (EPRI). The motivation for the R&D roadmap comes from the need to address the aging management of in-containment cables at nuclear power plants (NPPs).

  11. Light Water Reactor Sustainability (LWRS) Program – Non-Destructive Evaluation (NDE) R&D Roadmap for Determining Remaining Useful Life of Aging Cables in Nuclear Power Plants

    SciTech Connect (OSTI)

    Simmons, K.L.; Ramuhali, P.; Brenchley, D.L.; Coble, J.B.; Hashemian, H.M.; Konnick, R.; Ray, S.

    2012-09-01

    Executive Summary [partial] The purpose of the non-destructive evaluation (NDE) R&D Roadmap for Cables is to support the Materials Aging and Degradation (MAaD) R&D pathway. A workshop was held to gather subject matter experts to develop the NDE R&D Roadmap for Cables. The focus of the workshop was to identify the technical gaps in detecting aging cables and predicting their remaining life expectancy. The workshop was held in Knoxville, Tennessee, on July 30, 2012, at Analysis and Measurement Services Corporation (AMS) headquarters. The workshop was attended by 30 experts in materials, electrical engineering, and NDE instrumentation development from the U.S. Nuclear Regulatory Commission (NRC), U.S. Department of Energy (DOE) National Laboratories (Oak Ridge National Laboratory, Pacific Northwest National Laboratory, Argonne National Laboratory, and Idaho National Engineering Laboratory), universities, commercial NDE service vendors and cable manufacturers, and the Electric Power Research Institute (EPRI).

  12. Heat dissipating nuclear reactor

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Lazarus, Jonathan D. (Sunnyvale, CA)

    1987-01-01

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extends from the metal base plate downwardly and outwardly into the earth.

  13. Heat dissipating nuclear reactor

    DOE Patents [OSTI]

    Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

  14. Sustainabiliity Sustainability Plan

    E-Print Network [OSTI]

    Heller, Barbara

    Sustainabiliity IIT Campus Sustainability Plan 2010­2020 Published Fall Semester 2010 www.iit.edu/campus_sustainability #12;IIT Campus Sustainability Plan 2010-2020 Fall Semester 2010 1 Section I: Background Sustainability ...................................................................................................................................................................8 IIT Academic Entities on Sustainability

  15. SUSTAINABLE TRANSPORTATION ENERGY PATHWAYS

    E-Print Network [OSTI]

    California at Davis, University of

    SUSTAINABLE TRANSPORTATION ENERGY PATHWAYS A Research Summary for Decision Makers Edited by Joan Ogden and Lorraine Anderson #12;SUSTAINABLE TRANSPORTATION ENERGY PATHWAYS #12;SUSTAINABLE;6 SUSTAINABLE TRANSPORTATION ENERGY PATHWAYS ACKNOWLEDGEMENTS #12;1 SUSTAINABLE TRANSPORTATION ENERGY PATHWAYS

  16. Recommendations for Sustaining Sustainability Initiatives in City Government

    E-Print Network [OSTI]

    Boxumm-Debolt, Krystal; Edson, Greg; Holtzman, Karina; Jackson, Melissa; McCurdy, Nancy; Miller, David; Profaizer, Richard; Reinert, Doug; Wilburn, Sonta; Williams, Carla

    2012-08-15

    Conclusion 22 Appendix A: Denver, Colorado 23 Appendix B: New York City, New York 33 Appendix C: San Francisco, California 45 Appendix D: Seattle, Washington 53 Bibliography 62... and some of their sustainability program areas of excellence are: ? Denver, Colorado: Water, Energy, Environmental Governance ? New York City, New York: Waste, Air Quality, Water ? San Francisco, California: Waste, Air Quality, Green Buildings ? Seattle...

  17. Renewability and sustainability aspects of nuclear energy

    SciTech Connect (OSTI)

    ?ahin, Sümer

    2014-09-30

    Renewability and sustainability aspects of nuclear energy have been presented on the basis of two different technologies: (1) Conventional nuclear technology; CANDU reactors. (2) Emerging nuclear technology; fusion/fission (hybrid) reactors. Reactor grade (RG) plutonium, {sup 233}U fuels and heavy water moderator have given a good combination with respect to neutron economy so that mixed fuel made of (ThO{sub 2}/RG?PuO{sub 2}) or (ThC/RG-PuC) has lead to very high burn up grades. Five different mixed fuel have been selected for CANDU reactors composed of 4 % RG?PuO{sub 2} + 96 % ThO{sub 2}; 6 % RG?PuO{sub 2} + 94 % ThO{sub 2}; 10 % RG?PuO{sub 2} + 90 % ThO{sub 2}; 20 % RG?PuO{sub 2} + 80 % ThO{sub 2}; 30 % RG?PuO{sub 2} + 70 % ThO{sub 2}, uniformly taken in each fuel rod in a fuel channel. Corresponding operation lifetimes have been found as ? 0.65, 1.1, 1.9, 3.5, and 4.8 years and with burn ups of ? 30 000, 60 000, 100 000, 200 000 and 290 000 MW.d/ton, respectively. Increase of RG?PuO{sub 2} fraction in radial direction for the purpose of power flattening in the CANDU fuel bundle has driven the burn up grade to 580 000 MW.d/ton level. A laser fusion driver power of 500 MW{sub th} has been investigated to burn the minor actinides (MA) out of the nuclear waste of LWRs. MA have been homogenously dispersed as carbide fuel in form of TRISO particles with volume fractions of 0, 2, 3, 4 and 5 % in the Flibe coolant zone in the blanket surrounding the fusion chamber. Tritium breeding for a continuous operation of the fusion reactor is calculated as TBR = 1.134, 1.286, 1.387, 1.52 and 1.67, respectively. Fission reactions in the MA fuel under high energetic fusion neutrons have lead to the multiplication of the fusion energy by a factor of M = 3.3, 4.6, 6.15 and 8.1 with 2, 3, 4 and 5 % TRISO volume fraction at start up, respectively. Alternatively with thorium, the same fusion driver would produce ?160 kg {sup 233}U per year in addition to fission energy production in situ, multiplying the fusion energy by a factor of ?1.3.

  18. Operation of a steam hydro-gasifier in a fluidized bed reactor

    E-Print Network [OSTI]

    Park, Chan Seung; Norbeck, Joseph N.

    2008-01-01

    Using Self-Sustained Hydro- Gasification." [0011] In aprocess, using a steam hydro-gasification reactor (SHR) thepyrolysis and hydro-gasification in a single step. This

  19. Sustainability Framework 1 Queen's University

    E-Print Network [OSTI]

    Fletcher, Robin

    Sustainability Framework 1 Queen's University Sustainability Strategic Framework #12;Sustainability Framework 2 Contents Introduction .................................................................................................................................................. 2 Queen's Sustainability Mission

  20. Y-12 Site Sustainability Plan

    SciTech Connect (OSTI)

    Spencer, Charles G

    2012-12-01

    The accomplishments to date and the long-range planning of the Y-12 Energy Management and Sustainability and Stewardship programs support the U.S. Department of Energy (DOE) and the National Nuclear Security Administration (NNSA) vision for a commitment to energy effi ciency and sustainability and to achievement of the Guiding Principles. Specifi cally, the Y-12 vision is to support the Environment, Safety and Health Policy and the DOE Strategic Sustainability Performance Plan, while promoting overall sustainability and reduction of greenhouse gas emissions. The mission of the Y-12 Energy Management program is to incorporate energy-effi cient technologies site-wide and to position Y-12 to meet NNSA energy requirement needs through 2025 and beyond. The plan addresses greenhouse gases, buildings, fleet management, water use, pollution prevention, waste reduction, sustainable acquisition, electronic stewardship and data centers, site innovation and government-wide support.

  1. Steam turbine: Alternative emergency drive for the secure removal of residual heat from the core of light water reactors in ultimate emergency situation

    SciTech Connect (OSTI)

    Souza Dos Santos, R.

    2012-07-01

    In 2011 the nuclear power generation has suffered an extreme probation. That could be the meaning of what happened in Fukushima Nuclear Power Plants. In those plants, an earthquake of 8.9 on the Richter scale was recorded. The quake intensity was above the trip point of shutting down the plants. Since heat still continued to be generated, the procedure to cooling the reactor was started. One hour after the earthquake, a tsunami rocked the Fukushima shore, degrading all cooling system of plants. Since the earthquake time, the plant had lost external electricity, impacting the pumping working, drive by electric engine. When operable, the BWR plants responded the management of steam. However, the lack of electricity had degraded the plant maneuvers. In this paper we have presented a scheme to use the steam as an alternative drive to maintain operable the cooling system of nuclear power plant. This scheme adds more reliability and robustness to the cooling systems. Additionally, we purposed a solution to the cooling in case of lacking water for the condenser system. In our approach, steam driven turbines substitute electric engines in the ultimate emergency cooling system. (authors)

  2. Sustainable Supply Chain Networks Sustainable Cities

    E-Print Network [OSTI]

    Nagurney, Anna

    Chain Networks for Sustainable Cities #12;Pollution and Environmental Impacts In the US aloneSustainable Supply Chain Networks for Sustainable Cities Anna Nagurney John F. Smith Memorial Nagurney Sustainable Supply Chain Networks for Sustainable Cities #12;Acknowledgments I would like to thank

  3. Review: Hijacking Sustainability

    E-Print Network [OSTI]

    Antonelli, Monika

    2010-01-01

    addressing the attack on sustainability, Sharon Beder’s 2002Review: Hijacking Sustainability By Adrian Parr Reviewed byParr, Adrian. Hijacking Sustainability. Cambridge, MA: MIT

  4. Sustainability and Transport

    E-Print Network [OSTI]

    Gilbert, Richard

    2006-01-01

    2005. Integrating Sustainability into the Trans- portationTHOUGHT PIECE Sustainability and Transport by Richardof the concept of sustainability to transport planning. In

  5. Ecosystems and Sustainable Development

    E-Print Network [OSTI]

    Tufford, Dan

    1999-01-01

    Ecosystems and Sustainable Development Editors: J.L. Uso,Ecosystems and Sustainable Development. Southhampton, UK:ISBN: 1-85312-502-4. Sustainable development research is a

  6. Psychology of Sustainable Development

    E-Print Network [OSTI]

    Milfont, Taciano Lemos

    2003-01-01

    to achieve sustainable development: economic, environmental,Psychology of Sustainable Development By Peter Schmuck andPsychology of Sustainable Development. Norwell, MA: Kluwer

  7. Active stewardship: sustainable future

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

    Active stewardship: sustainable future Active stewardship: sustainable future Energy sustainability is a daunting task: How do we develop top-notch innovations with some of the...

  8. Sustainability | Department of Energy

    Office of Environmental Management (EM)

    Sustainability Sustainability The Department of Energy (DOE) supports the understanding and implementation of sustainability programs throughout the complex The Department of...

  9. www.planetearth.nerc.ac.uk Autumn 2014 Trout in hot water Biodiversity and big data Bioenergy's carbon footprint Sustainable drainage

    E-Print Network [OSTI]

    Brierley, Andrew

    's carbon footprint · Sustainable drainage Intothe #12;Front cover image courtesy Ben Langford About us NERC to account ­ bioenergy's carbon footprint What's the true cost of growing our fuel? 22 The science

  10. Water-Gas-Shift Membrane Reactor for High-Pressure Hydrogen Production. A comprehensive project report (FY2010 - FY2012)

    SciTech Connect (OSTI)

    Klaehn, John; Peterson, Eric; Orme, Christopher; Bhandari, Dhaval; Miller, Scott; Ku, Anthony; Polishchuk, Kimberly; Narang, Kristi; Singh, Surinder; Wei, Wei; Shisler, Roger; Wickersham, Paul; McEvoy, Kevin; Alberts, William; Howson, Paul; Barton, Thomas; Sethi, Vijay

    2013-01-01

    Idaho National Laboratory (INL), GE Global Research (GEGR), and Western Research Institute (WRI) have successfully produced hydrogen-selective membranes for water-gas-shift (WGS) modules that enable high-pressure hydrogen product streams. Several high performance (HP) polymer membranes were investigated for their gas separation performance under simulated (mixed gas) and actual syngas conditions. To enable optimal module performance, membranes with high hydrogen (H2) selectivity, permeance, and stability under WGS conditions are required. The team determined that the VTEC PI 80-051 and VTEC PI 1388 (polyimide from Richard Blaine International, Inc.) are prime candidates for the H2 gas separations at operating temperatures (~200°C). VTEC PI 80-051 was thoroughly analyzed for its H2 separations under syngas processing conditions using more-complex membrane configurations, such as tube modules and hollow fibers. These membrane formats have demonstrated that the selected VTEC membrane is capable of providing highly selective H2/CO2 separation (? = 7-9) and H2/CO separation (? = 40-80) in humidified syngas streams. In addition, the VTEC polymer membranes are resilient within the syngas environment (WRI coal gasification) at 200°C for over 1000 hours. The information within this report conveys current developments of VTEC PI 80-051 as an effective H2 gas separations membrane for high-temperature syngas streams.

  11. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Nuclear Energy Research Initiative Project 2001-001, Westinghouse Electric Co. Grant Number: DE-FG07-02SF22533, Final Report

    SciTech Connect (OSTI)

    Philip E. MacDonald

    2005-01-01

    The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% versus about 33% efficiency for current Light Water Reactors [LWRs]) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus, the need for a pressurizer, steam generators, steam separators, and dryers is eliminated. The main mission of the SCWR is generation of low-cost electricity. It is built upon two proven technologies: LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which are also in use around the world. The reference SCWR design for the U.S. program is a direct cycle system operating at 25.0 MPa, with core inlet and outlet temperatures of 280 and 500 C, respectively. The coolant density decreases from about 760 kg/m3 at the core inlet to about 90 kg/m3 at the core outlet. The inlet flow splits with about 10% of the inlet flow going down the space between the core barrel and the reactor pressure vessel (the downcomer) and about 90% of the inlet flow going to the plenum at the top of the rector pressure vessel, to then flow down through the core in special water rods to the inlet plenum. Here it mixes with the feedwater from the downcomer and flows upward to remove the heat in the fuel channels. This strategy is employed to provide good moderation at the top of the core. The coolant is heated to about 500 C and delivered to the turbine. The purpose of this NERI project was to assess the reference U.S. Generation IV SCWR design and explore alternatives to determine feasibility. The project was organized into three tasks: Task 1. Fuel-cycle Neutronic Analysis and Reactor Core Design Task 2. Fuel Cladding and Structural Material Corrosion and Stress Corrosion Cracking Task 3. Plant Engineering and Reactor Safety Analysis. moderator rods. materials.

  12. Design of Sustainable Supply Chains Sustainable Cities

    E-Print Network [OSTI]

    Nagurney, Anna

    Design of Sustainable Supply Chains for Sustainable Cities Anna Nagurney Isenberg School of Management University of Massachusetts Amherst, Massachusetts 01003 Complex-City Workshop Amsterdam.? Anna Nagurney Sustainability #12;Pollution and Environmental Impacts The release of CO2

  13. Moving from Aquifer Stress to Sustainable Management with Remote Sensing and Local Knowledge

    E-Print Network [OSTI]

    Richey, Alexandra S.

    2015-01-01

    Valley (CA): Orange County Water District. [cited 2015 SeptSustainability of ground-water resources: U.S. Geologicaltemporal allocation of ground water: a simulation approach.

  14. Sustainability Published December 2009

    E-Print Network [OSTI]

    Bandettini, Peter A.

    Sustainability Report Published December 2009 #12;#12;NIEHS Sustainability Report #12;#12;Table........................................................................................................9 Managing for Sustainability....................................................................................................45 #12;A Message from Our Director In order to sustain and improve life on Earth, we must

  15. Compact Reactor

    SciTech Connect (OSTI)

    Williams, Pharis E. [Williams Research, P.O. Box 554, Los Alamos, NM87544 (United States)

    2007-01-30

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date.

  16. Sustainable NREL

    ScienceCinema (OSTI)

    None

    2013-05-29

    The National Renewable Energy Laboratory prides itself on not only advancing the renewable energy, but "walking the talk" when it comes to sustainable practices. "When you look at our laboratories, you will see energy efficiency in action, but you'll also see renewable energy. We walk the walk and we talk the talk. We believe in it and we want to live it also."

  17. DOE Sustainability SPOtlight

    Broader source: Energy.gov [DOE]

    Newsletter highlights the recipients of the U.S. Department of Energy (DOE) Sustainability Performance Office (SPO) 2014 Sustainability Awards.

  18. Nuclear reactor vessel fuel thermal insulating barrier

    DOE Patents [OSTI]

    Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

    2013-03-19

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

  19. Track 4: Sustainability in National Programs I. Impact of Executive Orders on Federal Agencies

    E-Print Network [OSTI]

    Stoltz, Brian M.

    , water conservation, toxics reduction, high-performance buildings, electronics stewardship, sustainable of NASA's culture.!!! Department of Energy's Strategic Sustainability Performance Plan Abstract Josh1 Track 4: Sustainability in National Programs I. Impact of Executive Orders on Federal Agencies II

  20. MIT PEBBLE BED REACTOR PROJECT ANDREW C. KADAK

    E-Print Network [OSTI]

    , including the AP-600, the advanced boiling water reactor (ABWR) and the standard PWR designs offered of existing light water reactors and plans for evolutionary plants that were being considered at the timeMIT PEBBLE BED REACTOR PROJECT ANDREW C. KADAK Nuclear Science and Engineering Department