Sample records for modular nuclear reactors

  1. Small Modular Nuclear Reactors | 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: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What'sis Taking Over Our Instagram Secretary Moniz9MorganYou are here Home » SmallEnergyReactor

  2. Nuclear Safeguards Considerations For The Pebble Bed Modular Reactor (PBMR)

    SciTech Connect (OSTI)

    Phillip Casey Durst; David Beddingfield; Brian Boyer; Robert Bean; Michael Collins; Michael Ehinger; David Hanks; David L. Moses; Lee Refalo

    2009-10-01T23:59:59.000Z

    High temperature reactors (HTRs) have been considered since the 1940s, and have been constructed and demonstrated in the United Kingdom (Dragon), United States (Peach Bottom and Fort Saint Vrain), Japan (HTTR), Germany (AVR and THTR-300), and have been the subject of conceptual studies in Russia (VGM). The attraction to these reactors is that they can use a variety of reactor fuels, including abundant thorium, which upon reprocessing of the spent fuel can produce fissile U-233. Hence, they could extend the stocks of available uranium, provided the fuel is reprocessed. Another attractive attribute is that HTRs typically operate at a much higher temperature than conventional light water reactors (LWRs), because of the use of pyrolytic carbon and silicon carbide coated (TRISO) fuel particles embedded in ceramic graphite. Rather than simply discharge most of the unused heat from the working fluid in the power plant to the environment, engineers have been designing reactors for 40 years to recover this heat and make it available for district heating or chemical conversion plants. Demonstrating high-temperature nuclear energy conversion was the purpose behind Fort Saint Vrain in the United States, THTR-300 in Germany, HTTR in Japan, and HTR-10 and HTR-PM, being built in China. This resulted in nuclear reactors at least 30% or more thermodynamically efficient than conventional LWRs, especially if the waste heat can be effectively utilized in chemical processing plants. A modern variant of high temperature reactors is the Pebble Bed Modular Reactor (PBMR). Originally developed in the United States and Germany, it is now being redesigned and marketed by the Republic of South Africa and China. The team examined historical high temperature and high temperature gas reactors (HTR and HTGR) and reviewed safeguards considerations for this reactor. The following is a preliminary report on this topic prepared under the ASA-100 Advanced Safeguards Project in support of the NNSA Next Generation Safeguards Initiative (NGSI).

  3. Westinghouse Small Modular Reactor nuclear steam supply system design

    SciTech Connect (OSTI)

    Memmott, M. J.; Harkness, A. W.; Van Wyk, J. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

    2012-07-01T23:59:59.000Z

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam generator, and eight reactor coolant pumps (RCP). The containment vessel is 27.1 m (89 ft) long and 9.8 m (32 ft) in diameter, and is designed to withstand pressures up to 1.7 MPa (250 psi). It is completely submerged in a pool of water serving as a heat sink and radiation shield. Housed within the containment are four combined core makeup tanks (CMT)/passive residual heat removal (PRHR) heat exchangers, two in-containment pools (ICP), two ICP tanks and four valves which function as the automatic depressurization system (ADS). The PRHR heat exchangers are thermally connected to two different ultimate heat sink (UHS) tanks which provide transient cooling capabilities. (authors)

  4. SRS Small Modular Reactors

    SciTech Connect (OSTI)

    None

    2012-04-27T23:59:59.000Z

    The small modular reactor program at the Savannah River Site and the Savannah River National Laboratory.

  5. SRS Small Modular Reactors

    ScienceCinema (OSTI)

    None

    2014-05-21T23:59:59.000Z

    The small modular reactor program at the Savannah River Site and the Savannah River National Laboratory.

  6. ASSESSMENT OF SMALL AND MODULAR REACTOR NUCLEAR FUEL COST 

    E-Print Network [OSTI]

    Pannier, Christopher 1992-

    2012-05-03T23:59:59.000Z

    The nuclear energy industry is experiencing a renaissance of new reactor design and construction in Asia, North America, and Europe. The new Generation III designs are some of the largest ever built, featuring improved efficiency, construction...

  7. Multi-unit Operations in Non-Nuclear Systems: Lessons Learned for Small Modular Reactors

    SciTech Connect (OSTI)

    OHara J. M.; Higgins, J.; DAgostino, A.

    2012-01-17T23:59:59.000Z

    The nuclear-power community has reached the stage of proposing advanced reactor designs to support power generation for decades to come. Small modular reactors (SMRs) are one approach to meet these energy needs. While the power output of individual reactor modules is relatively small, they can be grouped to produce reactor sites with different outputs. Also, they can be designed to generate hydrogen, or to process heat. Many characteristics of SMRs are quite different from those of current plants and may be operated quite differently. One difference is that multiple units may be operated by a single crew (or a single operator) from one control room. The U.S. Nuclear Regulatory Commission (NRC) is examining the human factors engineering (HFE) aspects of SMRs to support licensing reviews. While we reviewed information on SMR designs to obtain information, the designs are not completed and all of the design and operational information is not yet available. Nor is there information on multi-unit operations as envisioned for SMRs available in operating experience. Thus, to gain a better understanding of multi-unit operations we sought the lesson learned from non-nuclear systems that have experience in multi-unit operations, specifically refineries, unmanned aerial vehicles and tele-intensive care units. In this paper we report the lessons learned from these systems and the implications for SMRs.

  8. Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 39, No. 11, p. 11691181 (November 2002) Conceptual Design of a Modular Island Core Fast Breeder Reactor "RAPID-M"

    E-Print Network [OSTI]

    Laughlin, Robert B.

    Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 39, No. 11, p. 1169­1181 (November 2002) Conceptual Design of a Modular Island Core Fast Breeder Reactor "RAPID-M" Mitsuru KAMBE Central Research Institute and accepted September 10, 2002) A metal fueled modular island core sodium cooled fast breeder reactor concept

  9. Computational Analysis of Fluid Flow in Pebble Bed Modular Reactor

    E-Print Network [OSTI]

    Gandhir, Akshay

    2012-10-19T23:59:59.000Z

    High Temperature Gas-cooled Reactor (HTGR) is a Generation IV reactor under consideration by Department of Energy and in the nuclear industry. There are two categories of HTGRs, namely, Pebble Bed Modular Reactor (PBMR) and Prismatic reactor. Pebble...

  10. Modular Inspection System for a Complete IN-Service Examination of Nuclear Reactor Pressure Vessel, Including Beltline Region

    SciTech Connect (OSTI)

    David H. Bothell

    2000-04-30T23:59:59.000Z

    Final Report for a DOE Phase II Contract Describing the design and fabrication of a reactor inspection modular rover prototype for reactor vessel inspection.

  11. Small Modular Reactors - SRSCRO

    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: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOnItemResearch > TheNuclearHomelandMultivariateSite Map Main Menu Aboutsmr Small Modular

  12. Design modification for the modular helium reactor for higher temperature operation and reliability studies for nuclear hydrogen production processes

    E-Print Network [OSTI]

    Reza, S.M. Mohsin

    2009-05-15T23:59:59.000Z

    Design options have been evaluated for the Modular Helium Reactor (MHR) for higher temperature operation. An alternative configuration for the MHR coolant inlet flow path is developed to reduce the peak vessel temperature (PVT). The coolant inlet...

  13. Advanced Modularity Design for The MIT Pebble Bed Reactor

    E-Print Network [OSTI]

    Advanced Modularity Design for The MIT Pebble Bed Reactor Andrew C. Kadak Department of Nuclear Reactor Technology Institute of Nuclear and New Energy Technology Friendship Hotel, Haidian District Beijing, China September 22-24, 2004 Abstract The future of all reactors will depend on whether they can

  14. Small Modular Reactors: Institutional Assessment

    SciTech Connect (OSTI)

    Joseph Perkowski, Ph.D.

    2012-06-01T23:59:59.000Z

    ? Objectives include, among others, a description of the basic development status of “small modular reactors” (SMRs) focused primarily on domestic activity; investigation of the domestic market appeal of modular reactors from the viewpoints of both key energy sector customers and also key stakeholders in the financial community; and consideration of how to proceed further with a pro-active "core group" of stakeholders substantially interested in modular nuclear deployment in order to provide the basis to expedite design/construction activity and regulatory approval. ? Information gathering was via available resources, both published and personal communications with key individual stakeholders; published information is limited to that already in public domain (no confidentiality); viewpoints from interviews are incorporated within. Discussions at both government-hosted and private-hosted SMR meetings are reflected herein. INL itself maintains a neutral view on all issues described. Note: as per prior discussion between INL and CAP, individual and highly knowledgeable senior-level stakeholders provided the bulk of insights herein, and the results of those interviews are the main source of the observations of this report. ? Attachment A is the list of individual stakeholders consulted to date, including some who provided significant earlier assessments of SMR institutional feasibility. ? Attachments B, C, and D are included to provide substantial context on the international status of SMR development; they are not intended to be comprehensive and are individualized due to the separate nature of the source materials. Attachment E is a summary of the DOE requirements for winning teams regarding the current SMR solicitation. Attachment F deserves separate consideration due to the relative maturity of the SMART SMR program underway in Korea. Attachment G provides illustrative SMR design features and is intended for background. Attachment H is included for overview purposes and is a sampling of advanced SMR concepts, which will be considered as part of the current DOE SMR program but whose estimated deployment time is beyond CAP’s current investment time horizon. Attachment I is the public DOE statement describing the present approach of their SMR Program.

  15. Advancing Small Modular Reactors: How We're Supporting Next-Gen...

    Energy Savers [EERE]

    Nuclear Energy Technology Advancing Small Modular Reactors: How We're Supporting Next-Gen Nuclear Energy Technology December 12, 2013 - 4:00pm Addthis The basics of small modular...

  16. Generic small modular reactor plant design.

    SciTech Connect (OSTI)

    Lewis, Tom Goslee,; Cipiti, Benjamin B.; Jordan, Sabina Erteza; Baum, Gregory A.

    2012-12-01T23:59:59.000Z

    This report gives an overview of expected design characteristics, concepts, and procedures for small modular reactors. The purpose of this report is to provide those who are interested in reducing the cost and improving the safety of advanced nuclear power plants with a generic design that possesses enough detail in a non-sensitive manner to give merit to their conclusions. The report is focused on light water reactor technology, but does add details on what could be different in a more advanced design (see Appendix). Numerous reactor and facility concepts were used for inspiration (documented in the bibliography). The final design described here is conceptual and does not reflect any proposed concept or sub-systems, thus any details given here are only relevant within this report. This report does not include any design or engineering calculations.

  17. Used nuclear fuel storage options including implications of small modular reactors

    E-Print Network [OSTI]

    Brinton, Samuel O. (Samuel Otis)

    2014-01-01T23:59:59.000Z

    This work addresses two aspects of the nuclear fuel cycle system with significant policy implications. The first is the preferred option for used fuel storage based on economics: local, regional or national storage. The ...

  18. Advancing Small Modular Reactors: How We're Supporting Next-Gen Nuclear

    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: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33Frequently20,000 Russian NuclearandJune 17, 2015EnergyTheAdvancedReactorEnergy Technology |

  19. Hybrid energy systems (HESs) using small modular reactors (SMRs)

    SciTech Connect (OSTI)

    S. Bragg-Sitton

    2014-10-01T23:59:59.000Z

    Large-scale nuclear reactors are traditionally operated for a singular purpose: steady-state production of dispatchable baseload electricity that is distributed broadly on the electric grid. While this implementation is key to a sustainable, reliable energy grid, small modular reactors (SMRs) offer new opportunities for increased use of clean nuclear energy for both electric and thermal ap plications in more locations – while still accommodating the desire to support renewable production sources.

  20. Evaluation of the applicability of existing nuclear power plant regulatory requirements in the U.S. to advanced small modular reactors.

    SciTech Connect (OSTI)

    LaChance, Jeffrey L.; Wheeler, Timothy A.; Farnum, Cathy Ottinger; Middleton, Bobby D.; Jordan, Sabina Erteza; Duran, Felicia Angelica; Baum, Gregory A.

    2013-05-01T23:59:59.000Z

    The current wave of small modular reactor (SMR) designs all have the goal of reducing the cost of management and operations. By optimizing the system, the goal is to make these power plants safer, cheaper to operate and maintain, and more secure. In particular, the reduction in plant staffing can result in significant cost savings. The introduction of advanced reactor designs and increased use of advanced automation technologies in existing nuclear power plants will likely change the roles, responsibilities, composition, and size of the crews required to control plant operations. Similarly, certain security staffing requirements for traditional operational nuclear power plants may not be appropriate or necessary for SMRs due to the simpler, safer and more automated design characteristics of SMRs. As a first step in a process to identify where regulatory requirements may be met with reduced staffing and therefore lower cost, this report identifies the regulatory requirements and associated guidance utilized in the licensing of existing reactors. The potential applicability of these regulations to advanced SMR designs is identified taking into account the unique features of these types of reactors.

  1. Proliferation resistance of small modular reactors fuels

    SciTech Connect (OSTI)

    Polidoro, F.; Parozzi, F. [RSE - Ricerca sul Sistema Energetico,Via Rubattino 54, 20134, Milano (Italy); Fassnacht, F.; Kuett, M.; Englert, M. [IANUS, Darmstadt University of Technology, Alexanderstr. 35, D-64283 Darmstadt (Germany)

    2013-07-01T23:59:59.000Z

    In this paper the proliferation resistance of different types of Small Modular Reactors (SMRs) has been examined and classified with criteria available in the literature. In the first part of the study, the level of proliferation attractiveness of traditional low-enriched UO{sub 2} and MOX fuels to be used in SMRs based on pressurized water technology has been analyzed. On the basis of numerical simulations both cores show significant proliferation risks. Although the MOX core is less proliferation prone in comparison to the UO{sub 2} core, it still can be highly attractive for diversion or undeclared production of nuclear material. In the second part of the paper, calculations to assess the proliferation attractiveness of fuel in typical small sodium cooled fast reactor show that proliferation risks from spent fuel cannot be neglected. The core contains a highly attractive plutonium composition during the whole life cycle. Despite some aspects of the design like the sealed core that enables easy detection of unauthorized withdrawal of fissile material and enhances proliferation resistance, in case of open Non-Proliferation Treaty break-out, weapon-grade plutonium in sufficient quantities could be extracted from the reactor core.

  2. Nuclear reactor engineering

    SciTech Connect (OSTI)

    Glasstone, S.; Sesonske, A.

    1981-01-01T23:59:59.000Z

    Chapters are presented concerning energy from nuclear fission; nuclear reactions and radiations; diffusion and slowing-down of neutrons; principles of reactor analysis; nuclear reactor kinetics and control; energy removal; non-fuel reactor materials; the reactor fuel system; radiation protection and environmental effects; nuclear reactor shielding; nuclear reactor safety; and power reactor systems.

  3. Sandia National Laboratories: Small Modular Reactors

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

    advanced computing capabilities that serve as a virtual version of existing, operating nuclear reactors-to enable nuclear energy to continue to provide dependable, afford-able...

  4. Design, analysis and optimization of the power conversion system for the Modular Pebble Bed Reactor System

    E-Print Network [OSTI]

    Wang, Chunyun, 1968-

    2003-01-01T23:59:59.000Z

    The Modular Pebble Bed Reactor system (MPBR) requires a gas turbine cycle (Brayton cycle) as the power conversion system for it to achieve economic competitiveness as a GenIV nuclear system. The availability of controllable ...

  5. The Modular Helium Reactor for Hydrogen Production

    SciTech Connect (OSTI)

    E. Harvego; M. Richards; A. Shenoy; K. Schultz; L. Brown; M. Fukuie

    2006-10-01T23:59:59.000Z

    For electricity and hydrogen production, an advanced reactor technology receiving considerable international interest is a modular, passively-safe version of the high-temperature, gas-cooled reactor (HTGR), known in the U.S. as the Modular Helium Reactor (MHR), which operates at a power level of 600 MW(t). For hydrogen production, the concept is referred to as the H2-MHR. Two concepts that make direct use of the MHR high-temperature process heat are being investigated in order to improve the efficiency and economics of hydrogen production. The first concept involves coupling the MHR to the Sulfur-Iodine (SI) thermochemical water splitting process and is referred to as the SI-Based H2-MHR. The second concept involves coupling the MHR to high-temperature electrolysis (HTE) and is referred to as the HTE-Based H2-MHR.

  6. Office of Nuclear Energy | Department of Energy

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

    Office of Nuclear Energy Small Modular Reactors Small Modular Reactors The Small Modular Reactor program advances the licensing and commercialization of this next-generation...

  7. ADVANCED SEISMIC BASE ISOLATION METHODS FOR MODULAR REACTORS

    SciTech Connect (OSTI)

    E. Blanford; E. Keldrauk; M. Laufer; M. Mieler; J. Wei; B. Stojadinovic; P.F. Peterson

    2010-09-20T23:59:59.000Z

    Advanced technologies for structural design and construction have the potential for major impact not only on nuclear power plant construction time and cost, but also on the design process and on the safety, security and reliability of next generation of nuclear power plants. In future Generation IV (Gen IV) reactors, structural and seismic design should be much more closely integrated with the design of nuclear and industrial safety systems, physical security systems, and international safeguards systems. Overall reliability will be increased, through the use of replaceable and modular equipment, and through design to facilitate on-line monitoring, in-service inspection, maintenance, replacement, and decommissioning. Economics will also receive high design priority, through integrated engineering efforts to optimize building arrangements to minimize building heights and footprints. Finally, the licensing approach will be transformed by becoming increasingly performance based and technology neutral, using best-estimate simulation methods with uncertainty and margin quantification. In this context, two structural engineering technologies, seismic base isolation and modular steel-plate/concrete composite structural walls, are investigated. These technologies have major potential to (1) enable standardized reactor designs to be deployed across a wider range of sites, (2) reduce the impact of uncertainties related to site-specific seismic conditions, and (3) alleviate reactor equipment qualification requirements. For Gen IV reactors the potential for deliberate crashes of large aircraft must also be considered in design. This report concludes that base-isolated structures should be decoupled from the reactor external event exclusion system. As an example, a scoping analysis is performed for a rectangular, decoupled external event shell designed as a grillage. This report also reviews modular construction technology, particularly steel-plate/concrete construction using factory prefabricated structural modules, for application to external event shell and base isolated structures.

  8. Nuclear reactor

    DOE Patents [OSTI]

    Thomson, Wallace B. (Severna Park, MD)

    2004-03-16T23:59:59.000Z

    A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

  9. Overview of the Westinghouse Small Modular Reactor building layout

    SciTech Connect (OSTI)

    Cronje, J. M. [Westinghouse Electric Company LLC, Centurion (South Africa); Van Wyk, J. J.; Memmott, M. J. [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

    2012-07-01T23:59:59.000Z

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the third in a series of four papers, which describe the design and functionality of the Westinghouse SMR. It focuses in particular upon the plant building layout and modular design of the Westinghouse SMR. In the development of small modular reactors, the building layout is an area where the safety of the plant can be improved by applying new design approaches. This paper will present an overview of the Westinghouse SMR building layout and indicate how the design features improve the safety and robustness of the plant. The Westinghouse SMR is designed with no shared systems between individual reactor units. The main buildings inside the security fence are the nuclear island, the rad-waste building, the annex building, and the turbine building. All safety related equipment is located in the nuclear island, which is a seismic class 1 building. To further enhance the safety and robustness of the design, the reactor, containment, and most of the safety related equipment are located below grade on the nuclear island. This reduces the possibility of severe damage from external threats or natural disasters. Two safety related ultimate heat sink (UHS) water tanks that are used for decay heat removal are located above grade, but are redundant and physically separated as far as possible for improved safety. The reactor and containment vessel are located below grade in the center of the nuclear island. The rad-waste and other radioactive systems are located on the bottom floors to limit the radiation exposure to personnel. The Westinghouse SMR safety trains are completely separated into four unconnected quadrants of the building, with access between quadrants only allowed above grade. This is an improvement to conventional reactor design since it prevents failures of multiple trains during floods or fires and other external events. The main control room is located below grade, with a remote shutdown room in a different quadrant. All defense in depth systems are placed on the nuclear island, primarily above grade, while the safety systems are located on lower floors. The economics of the Westinghouse SMR challenges the established approach of large Light Water Reactors (LWR) that utilized the economies of scale to reach economic competitiveness. To serve the market expectation of smaller capital investment and cost competitive energy, a modular design approach is implemented within the Westinghouse SMR. The Westinghouse SMR building layout integrates the three basic design constraints of modularization; transportation, handling and module-joining technology. (authors)

  10. Small Modular Reactors Presentation to Secretary of Energy Advisory...

    Office of Environmental Management (EM)

    of Energy Advisory Board - Deputy Assistant Secretary John Kelly DOE Small Modular Reactor Program (SMR) Research, Development & Deployment (RD&D) to enable the deployment of a...

  11. Human Reliability Analysis for Small Modular Reactors

    SciTech Connect (OSTI)

    Ronald L. Boring; David I. Gertman

    2012-06-01T23:59:59.000Z

    Because no human reliability analysis (HRA) method was specifically developed for small modular reactors (SMRs), the application of any current HRA method to SMRs represents tradeoffs. A first- generation HRA method like THERP provides clearly defined activity types, but these activity types do not map to the human-system interface or concept of operations confronting SMR operators. A second- generation HRA method like ATHEANA is flexible enough to be used for SMR applications, but there is currently insufficient guidance for the analyst, requiring considerably more first-of-a-kind analyses and extensive SMR expertise in order to complete a quality HRA. Although no current HRA method is optimized to SMRs, it is possible to use existing HRA methods to identify errors, incorporate them as human failure events in the probabilistic risk assessment (PRA), and quantify them. In this paper, we provided preliminary guidance to assist the human reliability analyst and reviewer in understanding how to apply current HRA methods to the domain of SMRs. While it is possible to perform a satisfactory HRA using existing HRA methods, ultimately it is desirable to formally incorporate SMR considerations into the methods. This may require the development of new HRA methods. More practicably, existing methods need to be adapted to incorporate SMRs. Such adaptations may take the form of guidance on the complex mapping between conventional light water reactors and small modular reactors. While many behaviors and activities are shared between current plants and SMRs, the methods must adapt if they are to perform a valid and accurate analysis of plant personnel performance in SMRs.

  12. Prognostics Health Management for Advanced Small Modular Reactor Passive Components

    SciTech Connect (OSTI)

    Meyer, Ryan M.; Ramuhalli, Pradeep; Coble, Jamie B.; Mitchell, Mark R.; Wootan, David W.; Hirt, Evelyn H.; Berglin, Eric J.; Bond, Leonard J.; Henager, Charles H.

    2013-10-18T23:59:59.000Z

    In the United States, sustainable nuclear power to promote energy security is a key national energy priority. Advanced small modular reactors (AdvSMR), which are based on modularization of advanced reactor concepts using non-light-water reactor (LWR) coolants such as liquid metal, helium, or liquid salt may provide a longer-term alternative to more conventional LWR-based concepts. The economics of AdvSMRs will be impacted by the reduced economy-of-scale savings when compared to traditional LWRs and the controllable day-to-day costs of AdvSMRs are expected to be dominated by operations and maintenance costs. Therefore, achieving the full benefits of AdvSMR deployment requires a new paradigm for plant design and management. In this context, prognostic health management of passive components in AdvSMRs can play a key role in enabling the economic deployment of AdvSMRs. In this paper, the background of AdvSMRs is discussed from which requirements for PHM systems are derived. The particle filter technique is proposed as a prognostics framework for AdvSMR passive components and the suitability of the particle filter technique is illustrated by using it to forecast thermal creep degradation using a physics-of-failure model and based on a combination of types of measurements conceived for passive AdvSMR components.

  13. Site Suitability and Hazard Assessment Guide for Small Modular Reactors

    SciTech Connect (OSTI)

    Wayne Moe

    2013-10-01T23:59:59.000Z

    Commercial nuclear reactor projects in the U.S. have traditionally employed large light water reactors (LWR) to generate regional supplies of electricity. Although large LWRs have consistently dominated commercial nuclear markets both domestically and abroad, the concept of small modular reactors (SMRs) capable of producing between 30 MW(t) and 900 MW(t) to generating steam for electricity is not new. Nor is the idea of locating small nuclear reactors in close proximity to and in physical connection with industrial processes to provide a long-term source of thermal energy. Growing problems associated continued use of fossil fuels and enhancements in efficiency and safety because of recent advancements in reactor technology suggest that the likelihood of near-term SMR technology(s) deployment at multiple locations within the United States is growing. Many different types of SMR technology are viable for siting in the domestic commercial energy market. However, the potential application of a particular proprietary SMR design will vary according to the target heat end-use application and the site upon which it is proposed to be located. Reactor heat applications most commonly referenced in connection with the SMR market include electric power production, district heating, desalinization, and the supply of thermal energy to various processes that require high temperature over long time periods, or a combination thereof. Indeed, the modular construction, reliability and long operational life purported to be associated with some SMR concepts now being discussed may offer flexibility and benefits no other technology can offer. Effective siting is one of the many early challenges that face a proposed SMR installation project. Site-specific factors dealing with support to facility construction and operation, risks to the plant and the surrounding area, and the consequences subsequent to those risks must be fully identified, analyzed, and possibly mitigated before a license will be granted to construct and operate a nuclear facility. Examples of significant site-related concerns include area geotechnical and geological hazard properties, local climatology and meteorology, water resource availability, the vulnerability of surrounding populations and the environmental to adverse effects in the unlikely event of radionuclide release, the socioeconomic impacts of SMR plant installation and the effects it has on aesthetics, proximity to energy use customers, the topography and area infrastructure that affect plant constructability and security, and concerns related to the transport, installation, operation and decommissioning of major plant components.

  14. Reactor & Nuclear Systems Publications | ORNL

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

    Science Home | Science & Discovery | Nuclear Science | Publications and Reports | Reactor and Nuclear Systems Publications SHARE Reactor and Nuclear Systems Publications The...

  15. PEBBLE-BED NUCLEAR REACTOR SYSTEM PHYSICS AND FUEL UTILIZATION

    E-Print Network [OSTI]

    Kelly, Ryan 1989-

    2011-04-20T23:59:59.000Z

    The Generation IV Pebble Bed Modular Reactor (PMBR) design may be used for electricity production, co-generation applications (industrial heat, hydrogen production, desalination, etc.), and could potentially eliminate some high level nuclear wastes...

  16. Baseline Concept Description of a Small Modular High Temperature Reactor

    SciTech Connect (OSTI)

    Hans Gougar

    2014-05-01T23:59:59.000Z

    The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the Generation IV program and its specific R&D needs will be included in this report when appropriate for comparison. The distinguishing features of the HTGR are the refractory (TRISO) coated particle fuel, the low-power density, graphite-moderated core, and the high outlet temperature of the inert helium coolant. The low power density and fuel form effectively eliminate the possibility of core melt, even upon a complete loss of coolant pressure and flow. The graphite, which constitutes the bulk of the core volume and mass, provides a large thermal buffer that absorbs fission heat such that thermal transients occur over a timespan of hours or even days. As chemically-inert helium is already a gas, there is no coolant temperature or void feedback on the neutronics and no phase change or corrosion product that could degrade heat transfer. Furthermore, the particle coatings and interstitial graphite retain fission products such that the source terms at the plant boundary remain well below actionable levels under all anticipated nominal and off-normal operating conditions. These attributes enable the reactor to supply process heat to a collocated industrial plant with negligible risk of contamination and minimal dynamic coupling of the facilities (Figure 1). The exceptional retentive properties of coated particle fuel in a graphite matrix were first demonstrated in the DRAGON reactor, a European research facility that began operation in 1964.

  17. Johnson Noise Thermometry for Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Britton, C.L.,Jr.; Roberts, M.; Bull, N.D.; Holcomb, D.E.; Wood, R.T.

    2012-09-15T23:59:59.000Z

    Temperature is a key process variable at any nuclear power plant (NPP). The harsh reactor environment causes all sensor properties to drift over time. At the higher temperatures of advanced NPPs the drift occurs more rapidly. The allowable reactor operating temperature must be reduced by the amount of the potential measurement error to assure adequate margin to material damage. Johnson noise is a fundamental expression of temperature and as such is immune to drift in a sensor’s physical condition. In and near the core, only Johnson noise thermometry (JNT) and radiation pyrometry offer the possibility for long-term, high-accuracy temperature measurement due to their fundamental natures. Small Modular Reactors (SMRs) place a higher value on long-term stability in their temperature measurements in that they produce less power per reactor core and thus cannot afford as much instrument recalibration labor as their larger brethren. The purpose of the current ORNL-led project, conducted under the Instrumentation, Controls, and Human-Machine Interface (ICHMI) research pathway of the U.S. Department of Energy (DOE) Advanced SMR Research and Development (R&D) program, is to develop and demonstrate a drift free Johnson noise-based thermometer suitable for deployment near core in advanced SMR plants.

  18. Johnson Noise Thermometry for Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Britton Jr, Charles L [ORNL; Roberts, Michael [ORNL; Bull, Nora D [ORNL; Holcomb, David Eugene [ORNL; Wood, Richard Thomas [ORNL

    2012-10-01T23:59:59.000Z

    Temperature is a key process variable at any nuclear power plant (NPP). The harsh reactor environment causes all sensor properties to drift over time. At the higher temperatures of advanced NPPs the drift occurs more rapidly. The allowable reactor operating temperature must be reduced by the amount of the potential measurement error to assure adequate margin to material damage. Johnson noise is a fundamental expression of temperature and as such is immune to drift in a sensor s physical condition. In and near core, only Johnson noise thermometry (JNT) and radiation pyrometry offer the possibility for long-term, high-accuracy temperature measurement due to their fundamental natures. Small, Modular Reactors (SMRs) place a higher value on long-term stability in their temperature measurements in that they produce less power per reactor core and thus cannot afford as much instrument recalibration labor as their larger brethren. The purpose of this project is to develop and demonstrate a drift free Johnson noise-based thermometer suitable for deployment near core in advanced SMR plants.

  19. Role of Nuclear Grade Graphite in Oxidation in Modular HTGRs

    SciTech Connect (OSTI)

    Willaim Windes; G. Strydom; J. Kane; R. Smith

    2014-11-01T23:59:59.000Z

    The passively safe High Temperature Gas-cooled Reactor (HTGR) design is one of the primary concepts considered for Generation IV and Small Modular Reactor (SMR) programs. The helium cooled, nuclear grade graphite moderated core achieves extremely high operating temperatures allowing either industrial process heat or electricity generation at high efficiencies. In addition to their neutron moderating properties, nuclear grade graphite core components provide excellent high temperature stability, thermal conductivity, and chemical compatibility with the high temperature nuclear fuel form. Graphite has been continuously used in nuclear reactors since the 1940’s and has performed remarkably well over a wide range of core environments and operating conditions. Graphite moderated, gas-cooled reactor designs have been safely used for research and power production purposes in multiple countries since the inception of nuclear energy development. However, graphite is a carbonaceous material, and this has generated a persistent concern that the graphite components could actually burn during either normal or accident conditions [ , ]. The common assumption is that graphite, since it is ostensibly similar to charcoal and coal, will burn in a similar manner. While charcoal and coal may have the appearance of graphite, the internal microstructure and impurities within these carbonaceous materials are very different. Volatile species and trapped moisture provide a source of oxygen within coal and charcoal allowing them to burn. The fabrication process used to produce nuclear grade graphite eliminates these oxidation enhancing impurities, creating a dense, highly ordered form of carbon possessing high thermal diffusivity and strongly (covalently) bonded atoms.

  20. Modularity in design of the MIT Pebble Bed Reactor

    E-Print Network [OSTI]

    Berte, Marc Vincent, 1977-

    2004-01-01T23:59:59.000Z

    The future of new nuclear power plant construction will depend in large part on the ability of designers to reduce capital, operations, and maintenance costs. One of the methods proposed, is to enhance the modularity of ...

  1. Reactor- Nuclear Science Center 

    E-Print Network [OSTI]

    Unknown

    2011-08-17T23:59:59.000Z

    A COMPARISON OF NUCLEAR REACTOR CONTROL ROOM DISPLAY PANELS A Thesis by FRANCES RENAE BOWERS Submitted to the Graduate College of Texas A&M University in partial fulfillment of the requirements for the degree of MASTER OF SCIENCE May 1988... Major Subject: Industrial Engineering A COMPARISON OF NUCLEAR REACTOR CONTROL ROOM DISPLAY PANELS A Thesis by FRANCES RENAE BOWERS Approved as to style and content by: Rod er . oppa (Cha' of 'ttee) R. Quinn Brackett (Member) rome . Co gleton...

  2. Development of a system model for advanced small modular reactors.

    SciTech Connect (OSTI)

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

    2014-01-01T23:59:59.000Z

    This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandia's concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

  3. ANALYSIS OF SEPCTRUM CHOICES FOR SMALL MODULAR REACTORS-PERFORMANCE AND DEVELOPMENT

    E-Print Network [OSTI]

    Kafle, Nischal

    2011-04-26T23:59:59.000Z

    the world; CAREM in Argentina, HTR-PM in China, FUJI in Japan, BREST, KLT series, SVBR-100, and VK-300 in Russia, PBMR (Pebble Bed Modular Reactor) in South Africa, SMART in South Korea, mPower, NuScale, GT- MHR in United States of America, and many... Association, 2011). 3 CAREM CAREM is a pressurized water reactor being studied in Argentina by INVAP (World Nuclear Association, 2010). This reactor uses 3.4% enriched uranium, and can produce 27 MWe which can be extendable to more than 300MWe...

  4. Analysis of granular flow in a pebble-bed nuclear reactor Chris H. Rycroft,1

    E-Print Network [OSTI]

    Bazant, Martin Z.

    Analysis of granular flow in a pebble-bed nuclear reactor Chris H. Rycroft,1 Gary S. Grest,2 James February 2006; published 24 August 2006 Pebble-bed nuclear reactor technology, which is currently being States, the Modular Pebble Bed Reactor MPBR 4,8 is a candidate for the next generation nuclear plant

  5. Population Sensitivity Evaluation of Two Proposed Hampton Roads Area Sites for a Possible Small Modular Reactor

    SciTech Connect (OSTI)

    Belles, R. J. [ORNL; Omitaomu, O. A. [ORNL

    2014-08-01T23:59:59.000Z

    The overall objective of this research project is to use the OR-SAGE tool to support the US Department of Energy (DOE) Office of Nuclear Energy (NE) in evaluating future electrical generation deployment options for small modular reactors (SMRs) in areas with significant energy demand from the federal sector. Deployment of SMRs in zones with high federal energy use can provide a means of meeting federal clean energy goals.

  6. Overall plant design specification Modular High Temperature Gas-cooled Reactor. Revision 9

    SciTech Connect (OSTI)

    NONE

    1990-05-01T23:59:59.000Z

    Revision 9 of the ``Overall Plant Design Specification Modular High Temperature Gas-Cooled Reactor,`` DOE-HTGR-86004 (OPDS) has been completed and is hereby distributed for use by the HTGR Program team members. This document, Revision 9 of the ``Overall Plant Design Specification`` (OPDS) reflects those changes in the MHTGR design requirements and configuration resulting form approved Design Change Proposals DCP BNI-003 and DCP BNI-004, involving the Nuclear Island Cooling and Spent Fuel Cooling Systems respectively.

  7. Nuclear reactor control

    SciTech Connect (OSTI)

    Ingham, R.V.

    1980-01-01T23:59:59.000Z

    A liquid metal cooled fast breeder nuclear reactor has power setback means for use in an emergency. On initiation of a trip-signal a control rod is injected into the core in two stages, firstly, by free fall to effect an immediate power-set back to a safe level and, secondly, by controlled insertion. Total shut-down of the reactor under all emergencies is avoided. 4 claims.

  8. Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Demonstration

    SciTech Connect (OSTI)

    Curtis Smith; Steven Prescott; Tony Koonce

    2014-04-01T23:59:59.000Z

    A key area of the Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) strategy is the development of methodologies and tools that will be used to predict the safety, security, safeguards, performance, and deployment viability of SMRs. The goal of the SMR PRA activity will be to develop quantitative methods and tools and the associated analysis framework for assessing a variety of risks. Development and implementation of SMR-focused safety assessment methods may require new analytic methods or adaptation of traditional methods to the advanced design and operational features of SMRs. We will need to move beyond the current limitations such as static, logic-based models in order to provide more integrated, scenario-based models based upon predictive modeling which are tied to causal factors. The development of SMR-specific safety models for margin determination will provide a safety case that describes potential accidents, design options (including postulated controls), and supports licensing activities by providing a technical basis for the safety envelope. This report documents the progress that was made to implement the PRA framework, specifically by way of demonstration of an advanced 3D approach to representing, quantifying and understanding flooding risks to a nuclear power plant.

  9. Nuclear reactor reflector

    DOE Patents [OSTI]

    Hopkins, Ronald J. (Pensacola, FL); Land, John T. (Pensacola, FL); Misvel, Michael C. (Pensacola, FL)

    1994-01-01T23:59:59.000Z

    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.

  10. Nuclear reactor reflector

    DOE Patents [OSTI]

    Hopkins, R.J.; Land, J.T.; Misvel, M.C.

    1994-06-07T23:59:59.000Z

    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.

  11. Nuclear reactor control column

    SciTech Connect (OSTI)

    Bachovchin, D.M.

    1982-08-10T23:59:59.000Z

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest crosssectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

  12. Nuclear reactor control column

    DOE Patents [OSTI]

    Bachovchin, Dennis M. (Plum Borough, PA)

    1982-01-01T23:59:59.000Z

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

  13. Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach

    SciTech Connect (OSTI)

    David Petti; Jim Kinsey; Dave Alberstein

    2014-01-01T23:59:59.000Z

    Various international efforts are underway to assess the safety of advanced nuclear reactor designs. For example, the International Atomic Energy Agency has recently held its first Consultancy Meeting on a new cooperative research program on high temperature gas-cooled reactor (HTGR) safety. Furthermore, the Generation IV International Forum Reactor Safety Working Group has recently developed a methodology, called the Integrated Safety Assessment Methodology, for use in Generation IV advanced reactor technology development, design, and design review. A risk and safety assessment white paper is under development with respect to the Very High Temperature Reactor to pilot the Integrated Safety Assessment Methodology and to demonstrate its validity and feasibility. To support such efforts, this information paper on the modular HTGR safety basis and approach has been prepared. The paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach. The paper gives those involved in the assessment of advanced reactor designs an opportunity to assess an advanced design that has already received extensive review by regulatory authorities and to judge the utility of recently proposed new methods for advanced reactor safety assessment such as the Integrated Safety Assessment Methodology.

  14. Nuclear Reactors and Technology

    SciTech Connect (OSTI)

    Cason, D.L.; Hicks, S.C. [eds.

    1992-01-01T23:59:59.000Z

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  15. Modularity of the MIT Pebble Bed Reactor for use by the commercial power industry

    E-Print Network [OSTI]

    Hanlon-Hyssong, Jaime E

    2008-01-01T23:59:59.000Z

    The Modular Pebble Bed Reactor is a small high temperature helium cooled reactor that is being considered for both electric power and hydrogen production. Pebble bed reactors are being developed in South Africa, China and ...

  16. Nuclear reactor safety device

    DOE Patents [OSTI]

    Hutter, Ernest (Wilmette, IL)

    1986-01-01T23:59:59.000Z

    A safety device is disclosed for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of an upward thermal excursion. Such safety device comprises a laminated helical ribbon configured as a tube-like helical coil having contiguous helical turns with slidably abutting edges. The helical coil is disclosed as a portion of a drive member connected axially to the control rod. The laminated ribbon is formed of outer and inner laminae. The material of the outer lamina has a greater thermal coefficient of expansion than the material of the inner lamina. In the event of an upward thermal excursion, the laminated helical coil curls inwardly to a smaller diameter. Such inward curling causes the total length of the helical coil to increase by a substantial increment, so that the control rod is axially repositioned by a corresponding amount to reduce the power output of the reactor.

  17. Heat dissipating nuclear reactor

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Lazarus, Jonathan D. (Sunnyvale, CA)

    1987-01-01T23:59:59.000Z

    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.

  18. Heat dissipating nuclear reactor

    DOE Patents [OSTI]

    Hunsbedt, A.; Lazarus, J.D.

    1985-11-21T23:59:59.000Z

    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.

  19. Modular Pebble Bed Reactor High Temperature Gas Reactor

    E-Print Network [OSTI]

    Built · Site Assembled · On--line Refueling · Modules added to meet demand. · No Reprocessing · High Experiments · Non-Proliferation · Safeguards · Waste Disposal · Reactor Research/ Demonstration Facility

  20. Nuclear reactor sealing system

    DOE Patents [OSTI]

    McEdwards, James A. (Calabasas, CA)

    1983-01-01T23:59:59.000Z

    A liquid metal-cooled nuclear reactor sealing system. The nuclear reactor includes a vessel sealed at its upper end by a closure head. The closure head comprises at least two components, one of which is rotatable; and the two components define an annulus therebetween. The sealing system includes at least a first and second inflatable seal disposed in series in an upper portion of the annulus. The system further includes a dip seal extending into a body of insulation located adjacent a bottom portion of the closure head. The dip seal comprises a trough formed by a lower portion of one of the components, and a seal blade pendently supported from the other component and extending downwardly into the trough. A body of liquid metal is contained in the trough which submerges a portion of the seal blade. The seal blade is provided with at least one aperture located above the body of liquid metal for providing fluid communication between the annulus intermediate the dip seal and the inflatable seals, and a body of cover gas located inside the vessel. There also is provided means for introducing a purge gas into the annulus intermediate the inflatable seals and the seal blade. The purge gas is introduced in an amount sufficient to substantially reduce diffusion of radioactive cover gas or sodium vapor up to the inflatable seals. The purge gas mixes with the cover gas in the reactor vessel where it can be withdrawn from the vessel for treatment and recycle to the vessel.

  1. Nuclear reactor building

    DOE Patents [OSTI]

    Gou, Perng-Fei (Saratoga, CA); Townsend, Harold E. (Campbell, CA); Barbanti, Giancarlo (Sirtori, IT)

    1994-01-01T23:59:59.000Z

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.

  2. Nuclear reactor building

    DOE Patents [OSTI]

    Gou, P.F.; Townsend, H.E.; Barbanti, G.

    1994-04-05T23:59:59.000Z

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed there above. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define there between an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin. 4 figures.

  3. Deep-Burn Modular Helium Reactor Fuel Development Plan

    SciTech Connect (OSTI)

    McEachern, D

    2002-12-02T23:59:59.000Z

    This document contains the workscope, schedule and cost for the technology development tasks needed to satisfy the fuel and fission product transport Design Data Needs (DDNs) for the Gas Turbine-Modular Helium Reactor (GT-MHR), operating in its role of transmuting transuranic (TRU) nuclides in spent fuel discharged from commercial light-water reactors (LWRs). In its application for transmutation, the GT-MHR is referred to as the Deep-Burn MHR (DB-MHR). This Fuel Development Plan (FDP) describes part of the overall program being undertaken by the U.S. Department of Energy (DOE), utilities, and industry to evaluate the use of the GT-MHR to transmute transuranic nuclides from spent nuclear fuel. The Fuel Development Plan (FDP) includes the work on fuel necessary to support the design and licensing of the DB-MHR. The FDP is organized into ten sections. Section 1 provides a summary of the most important features of the plan, including cost and schedule information. Section 2 describes the DB-MHR concept, the features of its fuel and the plan to develop coated particle fuel for transmutation. Section 3 describes the knowledge base for fabrication of coated particles, the experience with irradiation performance of coated particle fuels, the database for fission product transport in HTGR cores, and describes test data and calculations for the performance of coated particle fuel while in a repository. Section 4 presents the fuel performance requirements in terms of as-manufactured quality and performance of the fuel coatings under irradiation and accident conditions. These requirements are provisional because the design of the DB-MHR is in an early stage. However, the requirements are presented in this preliminary form to guide the initial work on the fuel development. Section 4 also presents limits on the irradiation conditions to which the coated particle fuel can be subjected for the core design. These limits are based on past irradiation experience. Section 5 describes the Design Data Needs to: (1) fabricate the coated particle fuel, (2) predict its performance in the reactor core, (3) predict the radionuclide release rates from the reactor core, and (4) predict the performance of spent fuel in a geological repository. The heart of this fuel development plan is Section 6, which describes the development activities proposed to satisfy the DDNs presented in Section 5. The development scope is divided into Fuel Process Development, Fuel Materials Development, Fission Product Transport, and Spent Fuel Disposal. Section 7 describes the facilities to be used. Generally, this program will utilize existing facilities. While some facilities will need to be modified, there is no requirement for major new facilities. Section 8 states the Quality Assurance requirements that will be applied to the development activities. Section 9 presents detailed costs organized by WBS and spread over time. Section 10 presents a list of the types of deliverables that will be prepared in each of the WBS elements. Four Appendices contain supplementary information on: (a) design data needs, (b) the interface with the separations plant, (c) the detailed development schedule, and (d) the detailed cost estimate.

  4. Nuclear Reactions and Reactor Safety

    E-Print Network [OSTI]

    Onuchic, José

    Nuclear Reactions and Reactor Safety DO NOT LICK We haven't entirely nailed down what element nuclear chain reaction, 1938 #12;Nuclear Chain Reactions Do nuclear chain reactions lead to runaway explosions? or ? -Controlled nuclear chain reactions possible: control energy release/sec -> More

  5. Nuclear reactor safety device

    DOE Patents [OSTI]

    Hutter, E.

    1983-08-15T23:59:59.000Z

    A safety device is described for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of a thermal excursion. It comprises a laminated strip helically configured to form a tube, said tube being in operative relation to said control rod. The laminated strip is formed of at least two materials having different thermal coefficients of expansion, and is helically configured such that the material forming the outer lamina of the tube has a greater thermal coefficient of expansion than the material forming the inner lamina of said tube. In the event of a thermal excursion the laminated strip will tend to curl inwardly so that said tube will increase in length, whereby as said tube increases in length it exerts a force on said control rod to axially reposition said control rod with respect to said core.

  6. Nuclear reactor control assembly

    SciTech Connect (OSTI)

    Negron, S.B.

    1991-06-11T23:59:59.000Z

    This patent describes an assembly for providing global power control in a nuclear reactor having the core split into two halves. It comprises a disk assembly formed from at least two disks each machined with an identical surface hole pattern such that rotation of one disk relative to the other causes the hole pattern to open or close, the disk assembly being positioned substantially at the longitudinal center of and coaxial with the core halves; and means for rotating at least one of the disks relative to the other.

  7. Nuclear reactor control apparatus

    SciTech Connect (OSTI)

    Sridhar, B.N.

    1981-08-28T23:59:59.000Z

    Nuclear reactor core safety rod release apparatus comprises a control rod having a detent notch in the form of an annular peripheral recess at its upper end, a control rod support tube for raising and lowering the control rod under normal conditions, latches pivotally mounted on the control support tube with free ends thereof normally disposed in the recess in the control rod, and cam means for pivoting the latches out of the recess in the control rod when a scram condition occurs. One embodiment of the invention comprises an additonal magnetically-operated latch for releasing the control rod under two different conditions, one involving seismic shock.

  8. Nuclear reactor control apparatus

    DOE Patents [OSTI]

    Sridhar, Bettadapur N. (Cupertino, CA)

    1983-11-01T23:59:59.000Z

    Nuclear reactor core safety rod release apparatus comprises a control rod having a detent notch in the form of an annular peripheral recess at its upper end, a control rod support tube for raising and lowering the control rod under normal conditions, latches pivotally mounted on the control support tube with free ends thereof normally disposed in the recess in the control rod, and cam means for pivoting the latches out of the recess in the control rod when a scram condition occurs. One embodiment of the invention comprises an additional magnetically-operated latch for releasing the control rod under two different conditions, one involving seismic shock.

  9. Pebble bed modular reactor safeguards: developing new approaches and implementing safeguards by design

    SciTech Connect (OSTI)

    Beyer, Brian David [Los Alamos National Laboratory; Beddingfield, David H [Los Alamos National Laboratory; Durst, Philip [INL; Bean, Robert [INL

    2010-01-01T23:59:59.000Z

    The design of the Pebble Bed Modular Reactor (PBMR) does not fit or seem appropriate to the IAEA safeguards approach under the categories of light water reactor (LWR), on-load refueled reactor (OLR, i.e. CANDU), or Other (prismatic HTGR) because the fuel is in a bulk form, rather than discrete items. Because the nuclear fuel is a collection of nuclear material inserted in tennis-ball sized spheres containing structural and moderating material and a PBMR core will contain a bulk load on the order of 500,000 spheres, it could be classified as a 'Bulk-Fuel Reactor.' Hence, the IAEA should develop unique safeguards criteria. In a multi-lab DOE study, it was found that an optimized blend of: (i) developing techniques to verify the plutonium content in spent fuel pebbles, (ii) improving burn-up computer codes for PBMR spent fuel to provide better understanding of the core and spent fuel makeup, and (iii) utilizing bulk verification techniques for PBMR spent fuel storage bins should be combined with the historic IAEA and South African approaches of containment and surveillance to verify and maintain continuity of knowledge of PBMR fuel. For all of these techniques to work the design of the reactor will need to accommodate safeguards and material accountancy measures to a far greater extent than has thus far been the case. The implementation of Safeguards-by-Design as the PBMR design progresses provides an approach to meets these safeguards and accountancy needs.

  10. Evaluation of the Gas Turbine Modular Helium Reactor

    SciTech Connect (OSTI)

    Not Available

    1994-02-01T23:59:59.000Z

    Recent advances in gas-turbine and heat exchanger technology have enhanced the potential for a Modular Helium Reactor (MHR) incorporating a direct gas turbine (Brayton) cycle for power conversion. The resulting Gas Turbine Modular Helium Reactor (GT-MHR) power plant combines the high temperature capabilities of the MHR with the efficiency and reliability of modern gas turbines. While the passive safety features of the steam cycle MHR (SC-MHR) are retained, generation efficiencies are projected to be in the range of 48% and steam power conversion systems, with their attendant complexities, are eliminated. Power costs are projected to be reduced by about 20%, relative to the SC-MHR or coal. This report documents the second, and final, phase of a two-part evaluation that concluded with a unanimous recommendation that the direct cycle (DC) variant of the GT-MHR be established as the commercial objective of the US Gas-Cooled Reactor Program. This recommendation has been endorsed by industrial and utility participants and accepted by the US Department of Energy (DOE). The Phase II effort, documented herein, concluded that the DC GT-MHR offers substantial technical and economic advantages over both the IDC and SC systems. Both the DC and IDC were found to offer safety advantages, relative to the SC, due to elimination of the potential for water ingress during power operations. This is the dominant consequence event for the SC. The IDC was judged to require somewhat less development than the direct cycle, while the SC, which has the greatest technology base, incurs the least development cost and risk. While the technical and licensing requirements for the DC were more demanding, they were judged to be incremental and feasible. Moreover, the DC offers significant performance and cost improvements over the other two concepts. Overall, the latter were found to justify the additional development needs.

  11. Nuclear reactor control

    DOE Patents [OSTI]

    Cawley, William E. (Phoenix, AZ); Warnick, Robert F. (Pasco, WA)

    1982-01-01T23:59:59.000Z

    1. In a nuclear reactor incorporating a plurality of columns of tubular fuel elements disposed in horizontal tubes in a mass of graphite wherein water flows through the tubes to cool the fuel elements, the improvement comprising at least one control column disposed in a horizontal tube including fewer fuel elements than in a normal column of fuel elements and tubular control elements disposed at both ends of said control column, and means for varying the horizontal displacement of the control column comprising a winch at the upstream end of the control column and a cable extending through the fuel and control elements and attached to the element at the downstream end of the column.

  12. Nuclear reactor control apparatus

    DOE Patents [OSTI]

    Sridhar, Bettadapur N. (Cupertino, CA)

    1983-10-25T23:59:59.000Z

    Nuclear reactor safety rod release apparatus comprises a ring which carries detents normally positioned in an annular recess in outer side of the rod, the ring being held against the lower end of a drive shaft by magnetic force exerted by a solenoid carried by the drive shaft. When the solenoid is de-energized, the detent-carrying ring drops until the detents contact a cam surface associated with the lower end of the drive shaft, at which point the detents are cammed out of the recess in the safety rod to release the rod from the drive shaft. In preferred embodiments of the invention, an additional latch is provided to release a lower portion of a safety rod under conditions that may interfere with movement of the entire rod.

  13. Nuclear reactor control

    SciTech Connect (OSTI)

    Cawley, W.E.; Warnick, R.F.

    1982-03-30T23:59:59.000Z

    In a nuclear reactor incorporating a plurality of columns of tubular fuel elements disposed in horizontal tubes in a mass of graphite wherein water flows through the tubes to cool the fuel elements, the improvement comprising at least one control column disposed in a horizontal tube including fewer fuel elements than in a normal column of fuel elements and tubular control elements disposed at both ends of said control column, and means for varying the horizontal displacement of the control column comprising a winch at the upstream end of the control column and a cable extending through the fuel and control elements and attached to the element at the downstream end of the column.

  14. Nuclear reactor engineering

    SciTech Connect (OSTI)

    Glasstone, S.; Sesonske, A.

    1982-07-01T23:59:59.000Z

    A book is reviewed which emphasizes topics directly related to the light water reactor power plant and the fast reactor power system. Current real-world problems are addressed throughout the text, and a chapter on safety includes much of the postThree Mile Island impact on operating systems. Topics covered include Doppler broadening, neutron resonances, multigroup diffusion theory, reactor kinetics, reactor control, energy removal, nonfuel materials, reactor fuel, radiation protection, environmental effects, and reactor safety.

  15. Development of the Mathematics of Learning Curve Models for Evaluating Small Modular Reactor Economics

    SciTech Connect (OSTI)

    Harrison, T. J. [ORNL

    2014-02-01T23:59:59.000Z

    The cost of nuclear power is a straightforward yet complicated topic. It is straightforward in that the cost of nuclear power is a function of the cost to build the nuclear power plant, the cost to operate and maintain it, and the cost to provide fuel for it. It is complicated in that some of those costs are not necessarily known, introducing uncertainty into the analysis. For large light water reactor (LWR)-based nuclear power plants, the uncertainty is mainly contained within the cost of construction. The typical costs of operations and maintenance (O&M), as well as fuel, are well known based on the current fleet of LWRs. However, the last currently operating reactor to come online was Watts Bar 1 in May 1996; thus, the expected construction costs for gigawatt (GW)-class reactors in the United States are based on information nearly two decades old. Extrapolating construction, O&M, and fuel costs from GW-class LWRs to LWR-based small modular reactors (SMRs) introduces even more complication. The per-installed-kilowatt construction costs for SMRs are likely to be higher than those for the GW-class reactors based on the property of the economy of scale. Generally speaking, the economy of scale is the tendency for overall costs to increase slower than the overall production capacity. For power plants, this means that doubling the power production capacity would be expected to cost less than twice as much. Applying this property in the opposite direction, halving the power production capacity would be expected to cost more than half as much. This can potentially make the SMRs less competitive in the electricity market against the GW-class reactors, as well as against other power sources such as natural gas and subsidized renewables. One factor that can potentially aid the SMRs in achieving economic competitiveness is an economy of numbers, as opposed to the economy of scale, associated with learning curves. The basic concept of the learning curve is that the more a new process is repeated, the more efficient the process can be made. Assuming that efficiency directly relates to cost means that the more a new process is repeated successfully and efficiently, the less costly the process can be made. This factor ties directly into the factory fabrication and modularization aspect of the SMR paradigm—manufacturing serial, standardized, identical components for use in nuclear power plants can allow the SMR industry to use the learning curves to predict and optimize deployment costs.

  16. Fast reactors and nuclear nonproliferation

    SciTech Connect (OSTI)

    Avrorin, E.N. [Russian Federal Nuclear Center - Zababakhin Institute of Applied Physics, Snezhinsk (Russian Federation); Rachkov, V.I.; Chebeskov, A.N. [State Scientific Center of the Russian Federation - Institute for Physics and Power Engineering, Bondarenko Square, 1, Obninsk, Kaluga region, 249033 (Russian Federation)

    2013-07-01T23:59:59.000Z

    Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (authors)

  17. An Overview of the Safety Case for Small Modular Reactors

    SciTech Connect (OSTI)

    Ingersoll, Daniel T [ORNL] [ORNL

    2011-01-01T23:59:59.000Z

    Several small modular reactor (SMR) designs emerged in the late 1970s and early 1980s in response to lessons learned from the many technical and operational challenges of the large Generation II light-water reactors. After the accident at the Three Mile Island plant in 1979, an ensuing reactor redesign effort spawned the term inherently safe designs, which later evolved into passively safe terminology. Several new designs were engineered to be deliberately small in order to fully exploit the benefits of passive safety. Today, new SMR designs are emerging with a similar philosophy of offering highly robust and resilient designs with increased safety margins. Additionally, because these contemporary designs are being developed subsequent to the September 11, 2001, terrorist attack, they incorporate a number of intrinsic design features to further strengthen their safety and security. Several SMR designs are being developed in the United States spanning the full spectrum of reactor technologies, including water-, gas-, and liquid-metal-cooled ones. Despite a number of design differences, most of these designs share a common set of design principles to enhance plant safety and robustness, such as eliminating plant design vulnerabilities where possible, reducing accident probabilities, and mitigating accident consequences. An important consequence of the added resilience provided by these design approaches is that the individual reactor units and the entire plant should be able to survive a broader range of extreme conditions. This will enable them to not only ensure the safety of the general public but also help protect the investment of the owner and continued availability of the power-generating asset. Examples of typical SMR design features and their implications for improved plant safety are given for specific SMR designs being developed in the United States.

  18. Human Factors Issues For Multi-Modular Reactor Units

    SciTech Connect (OSTI)

    Tuan Q Tran; Humberto E. Garcia; Ronald L. Boring; Jeffrey C. Joe; Bruce P. Hallbert

    2007-08-01T23:59:59.000Z

    Smaller and multi-modular reactor (MMR) will be highly technologically-advanced systems allowing more system flexibility to reactors configurations (e.g., addition/deletion of reactor units). While the technical and financial advantages of systems may be numerous, MMR presents many human factors challenges that may pose vulnerability to plant safety. An important human factors challenge in MMR operation and performance is the monitoring of data from multiple plants from centralized control rooms where human operators are responsible for interpreting, assessing, and responding to different system’s states and failures (e.g., simultaneously monitoring refueling at one plant while keeping an eye on another plant’s normal operating state). Furthermore, the operational, safety, and performance requirements for MMR can seriously change current staffing models and roles, the mode in which information is displayed, procedures and training to support and guide operators, and risk analysis. For these reasons, addressing human factors concerns in MMR are essential in reducing plant risk.

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

    E-Print Network [OSTI]

    Terrani, Kurt Amir

    2010-01-01T23:59:59.000Z

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

  20. The design of a reduced diameter Pebble Bed Modular Reactor for reactor pressure vessel transport by railcar

    E-Print Network [OSTI]

    Everson, Matthew S

    2009-01-01T23:59:59.000Z

    Many desirable locations for Pebble Bed Modular Reactors are currently out of consideration as construction sites for current designs due to limitations on the mode of transportation for large RPVs. In particular, the ...

  1. Teaching About Nature's Nuclear Reactors

    E-Print Network [OSTI]

    Herndon, J M

    2005-01-01T23:59:59.000Z

    Naturally occurring nuclear reactors existed in uranium deposits on Earth long before Enrico Fermi built the first man-made nuclear reactor beneath Staggs Field in 1942. In the story of their discovery, there are important lessons to be learned about scientific inquiry and scientific discovery. Now, there is evidence to suggest that the Earth's magnetic field and Jupiter's atmospheric turbulence are driven by planetary-scale nuclear reactors. The subject of planetocentric nuclear fission reactors can be a jumping off point for stimulating classroom discussions about the nature and implications of planetary energy sources and about the geomagnetic field. But more importantly, the subject can help to bring into focus the importance of discussing, debating, and challenging current thinking in a variety of areas.

  2. Nuclear Reactors and Technology; (USA)

    SciTech Connect (OSTI)

    Cason, D.L.; Hicks, S.C. (eds.)

    1991-01-01T23:59:59.000Z

    Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database (EDB) during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency's Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on EDB and Nuclear Science Abstracts (NSA) database. Current information, added daily to EDB, is available to DOE and its contractors through the DOE integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user's needs.

  3. Nuclear reactor control rod

    SciTech Connect (OSTI)

    Cearley, J.E.; Izzo, K.R.

    1987-06-30T23:59:59.000Z

    This patent describes a vertically oriented bottom entry control rod from a nuclear reactor: a frame including an elongated central spine of cruciform cross section connected between an upper support member and a lower support member both of cruciform shape having four laterally extending arms. The arms are in alignment with the arms of the lower support member and each aligned upper and lower support members has a sheath extending between; absorber plates of neutron absorber material, different from the material of the frame, one of the absorber plates is positioned within a sheath beneath each of the arms; attachment means suspends the absorber plates from the arms of the upper support member within a sheath; elongated absorber members positioned within a sheath between each of the suspended absorber plates and an arm of the lower support member; and joint means between the upper ends of the absorber members and the lower ends of the suspended absorber plates for minimizing gaps; the sheath means encloses the suspended absorber plates and the absorber members extending between aligned arms of the upper and lower support members and secured.

  4. Nuclear Reactor Safety Design Criteria

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1993-01-19T23:59:59.000Z

    The order establishes nuclear safety criteria applicable to the design, fabrication, construction, testing, and performance requirements of nuclear reactor facilities and safety class structures, systems, and components (SSCs) within these facilities. Cancels paragraphs 8a and 8b of DOE 5480.6. Cancels DOE O 5480.6 in part. Certified 11-18-10.

  5. Uncertainties in the Anti-neutrino Production at Nuclear Reactors

    E-Print Network [OSTI]

    Djurcic, Zelimir

    2009-01-01T23:59:59.000Z

    neutrino Production at Nuclear Reactors Z. Djurcic 1 , ?emission rates from nuclear reactors are determined fromlarge commercial nuclear reactors are playing an important

  6. A Framework to Expand and Advance Probabilistic Risk Assessment to Support Small Modular Reactors

    SciTech Connect (OSTI)

    Curtis Smith; David Schwieder; Robert Nourgaliev; Cherie Phelan; Diego Mandelli; Kellie Kvarfordt; Robert Youngblood

    2012-09-01T23:59:59.000Z

    During the early development of nuclear power plants, researchers and engineers focused on many aspects of plant operation, two of which were getting the newly-found technology to work and minimizing the likelihood of perceived accidents through redundancy and diversity. As time, and our experience, has progressed, the realization of plant operational risk/reliability has entered into the design, operation, and regulation of these plants. But, to date, we have only dabbled at the surface of risk and reliability technologies. For the next generation of small modular reactors (SMRs), it is imperative that these technologies evolve into an accepted, encompassing, validated, and integral part of the plant in order to reduce costs and to demonstrate safe operation. Further, while it is presumed that safety margins are substantial for proposed SMR designs, the depiction and demonstration of these margins needs to be better understood in order to optimize the licensing process.

  7. Design of an Alternative Coolant Inlet Flow Configuaration for the Modular Helium Reactor

    SciTech Connect (OSTI)

    SM Mohsin Reza; E. A. Harvego; Matt Richards; Arkal Shenoy; Kenneth Lee Peddicord

    2006-06-01T23:59:59.000Z

    The coolant outlet temperature for the Modular Helium Reactor (MHR) was increased to improve the overall efficiency of nuclear hydrogen production using either thermochemical or high temperature electrolysis (HTE) processes. The inlet temperature was also increased to keep about the same _T across the reactor core. Thermal hydraulic analyses of the current MHR design were performed with these updated temperatures to determine the impact of these highter temperatures on pressure drops, coolant flow rates and temperature profiles within the vessel and core regions. Due to these increased operating temperatures, the overall efficiency of hydrogen production processes increases but the steady state reactor vessel temperature is found to be well above the ASME code limits for current vessel materials. Using the RELAP5-3D/ATHENA computer code, an alternative configuration for the MHR coolant inlet flow path was evaluated in an attempt to reduce the reactor vessel temperatures. The coolant inlet flow was shifted from channel boxes located in the annular region between the reactor core barrel and the inner wall of the reactor vessel to a flow path through the outer permanent reflector. Considering the available thickness of graphite in the permanent outer reflector, the total flow area, the number of coolant holes and the coolant-hole diameter were varied to optimize the pressure drop, the coolant inlet velocity and the percentage of graphite removed from the core. The resulting thermal hydraulic analyses of the optimized design showed that peak vessel and fuel temperatures were within acceptable limits for both steady-state and transient operating conditions.

  8. Nuclear reactor downcomer flow deflector

    DOE Patents [OSTI]

    Gilmore, Charles B. (Greensburg, PA); Altman, David A. (Pittsburgh, PA); Singleton, Norman R. (Murrysville, PA)

    2011-02-15T23:59:59.000Z

    A nuclear reactor having a coolant flow deflector secured to a reactor core barrel in line with a coolant inlet nozzle. The flow deflector redirects incoming coolant down an annulus between the core barrel and the reactor vessel. The deflector has a main body with a front side facing the fluid inlet nozzle and a rear side facing the core barrel. The rear side of the main body has at least one protrusion secured to the core barrel so that a gap exists between the rear side of the main body adjacent the protrusion and the core barrel. Preferably, the protrusion is a relief that circumscribes the rear side of the main body.

  9. Nuclear reactor safety heat transfer

    SciTech Connect (OSTI)

    Jones, O.C.

    1982-07-01T23:59:59.000Z

    Reviewed is a book which has 5 parts: Overview, Fundamental Concepts, Design Basis Accident-Light Water Reactors (LWRs), Design Basis Accident-Liquid-Metal Fast Breeder Reactors (LMFBRs), and Special Topics. It combines a historical overview, textbook material, handbook information, and the editor's personal philosophy on safety of nuclear power plants. Topics include thermal-hydraulic considerations; transient response of LWRs and LMFBRs following initiating events; various accident scenarios; single- and two-phase flow; single- and two-phase heat transfer; nuclear systems safety modeling; startup and shutdown; transient response during normal and upset conditions; vapor explosions, natural convection cooling; blockages in LMFBR subassemblies; sodium boiling; and Three Mile Island.

  10. Utility/user requirements for the Modular High Temperature Gas-Cooled Reactor Plant

    SciTech Connect (OSTI)

    Swart, F.E.

    1987-06-01T23:59:59.000Z

    The purpose of this document is to set forth the top level Utilty/User requirements for a Modular High Temperature Gas-Cooled Reactor electric generating plant that incorporates 4 reactors and 2 turbine-generators to produce a nominal electrical output of 550 MW net.

  11. MIT Modular Pebble Bed Reactor (MPBR) A Summary of Research Activities and Accomplishments

    E-Print Network [OSTI]

    Disposal · Reactor Research/ Demonstration Facility · License by Test · Expert I&C System - Hands free.71MPa 69.7 C 4.67MPa Cooling RPV #12;BOP System Analysis and Dynamic Simulation Model DevelopmentMIT Modular Pebble Bed Reactor (MPBR) A Summary of Research Activities and Accomplishments Andrew C

  12. Interdisciplinary Institute for Innovation Nuclear reactors' construction

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Interdisciplinary Institute for Innovation Nuclear reactors' construction costs: The role of lead@mines-paristech.fr hal-00956292,version1-6Mar2014 #12;hal-00956292,version1-6Mar2014 #12;Nuclear reactors' construction reactor construction costs in France and the United States. Studying the cost of nuclear power has often

  13. Nuclear reactor engineering: Reactor design basics. Fourth edition, Volume One

    SciTech Connect (OSTI)

    Glasstone, S.; Sesonske, A.

    1994-12-31T23:59:59.000Z

    This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in design and operation of nuclear power plants. Extensively updated, the fourth edition includes new material on reactor safety and risk analysis, regulation, fuel management, waste management, and operational aspects of nuclear power. This volume contains the following: energy from nuclear fission; nuclear reactions and radiations; neutron transport; nuclear design basics; nuclear reactor kinetics and control; radiation protection and shielding; and reactor materials.

  14. Nuclear reactor engineering: Reactor systems engineering. Fourth edition, Volume Two

    SciTech Connect (OSTI)

    Glasstone, S.; Sesonske, A.

    1994-12-31T23:59:59.000Z

    This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in the design and operation of nuclear power plants. Extensively updated, the fourth edition includes new materials on reactor safety and risk analysis, regulation, fuel management, waste management and operational aspects of nuclear power. This volume contains the following: the systems concept, design decisions, and information tools; energy transport; reactor fuel management and energy cost considerations; environmental effects of nuclear power and waste management; nuclear reactor safety and regulation; power reactor systems; plant operations; and advanced plants and the future.

  15. Westinghouse Small Modular Reactor balance of plant and supporting systems design

    SciTech Connect (OSTI)

    Memmott, M. J.; Stansbury, C.; Taylor, C. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

    2012-07-01T23:59:59.000Z

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operation of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)

  16. 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)

    1992-01-01T23:59:59.000Z

    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.

  17. Incorporating Equipment Condition Assessment in Risk Monitors for Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Coble, Jamie B.; Coles, Garill A.; Meyer, Ryan M.; Ramuhalli, Pradeep

    2013-10-01T23:59:59.000Z

    Advanced small modular reactors (aSMRs) can complement the current fleet of large light-water reactors in the USA for baseload and peak demand power production and process heat applications (e.g., water desalination, shale oil extraction, hydrogen production). The day-to-day costs of aSMRs are expected to be dominated by operations and maintenance (O&M); however, the effect of diverse operating missions and unit modularity on O&M is not fully understood. These costs could potentially be reduced by optimized scheduling, with risk-informed scheduling of maintenance, repair, and replacement of equipment. Currently, most nuclear power plants have a “living” probabilistic risk assessment (PRA), which reflects the as-operated, as-modified plant and combine event probabilities with population-based probability of failure (POF) for key components. “Risk monitors” extend the PRA by incorporating the actual and dynamic plant configuration (equipment availability, operating regime, environmental conditions, etc.) into risk assessment. In fact, PRAs are more integrated into plant management in today’s nuclear power plants than at any other time in the history of nuclear power. However, population-based POF curves are still used to populate fault trees; this approach neglects the time-varying condition of equipment that is relied on during standard and non-standard configurations. Equipment condition monitoring techniques can be used to estimate the component POF. Incorporating this unit-specific estimate of POF in the risk monitor can provide a more accurate estimate of risk in different operating and maintenance configurations. This enhanced risk assessment will be especially important for aSMRs that have advanced component designs, which don’t have an available operating history to draw from, and often use passive design features, which present challenges to PRA. This paper presents the requirements and technical gaps for developing a framework to integrate unit-specific estimates of POF into risk monitors, resulting in enhanced risk monitors that support optimized operation and maintenance of aSMRs.

  18. NRC Reviewer Aid for Evaluating the Human Factors Engineering Aspects of Small Modular Reactors

    SciTech Connect (OSTI)

    OHara J. M.; Higgins, J.C.

    2012-01-13T23:59:59.000Z

    Small modular reactors (SMRs) are a promising approach to meeting future energy needs. Although the electrical output of an individual SMR is relatively small compared to that of typical commercial nuclear plants, they can be grouped to produce as much energy as a utility demands. Furthermore, SMRs can be used for other purposes, such as producing hydrogen and generating process heat. The design characteristics of many SMRs differ from those of current conventional plants and may require a distinct concept of operations (ConOps). The U.S. Nuclear Regulatory Commission (NRC) conducted research to examine the human factors engineering (HFE) and the operational aspects of SMRs. The research identified thirty potential human-performance issues that should be considered in the NRC's reviews of SMR designs and in future research activities. The purpose of this report is to support NRC HFE reviewers of SMR applications by identifying some of the questions that can be asked of applicants whose designs have characteristics identified in the issues. The questions for each issue were identified and organized based on the review elements and guidance contained in Chapter 18 of the Standard Review Plan (NUREG-0800), and the Human Factors Engineering Program Review Model (NUREG-0711).

  19. Technical Needs for Enhancing Risk Monitors with Equipment Condition Assessment for Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Coble, Jamie B.; Coles, Garill A.; Ramuhalli, Pradeep; Meyer, Ryan M.; Berglin, Eric J.; Wootan, David W.; Mitchell, Mark R.

    2013-04-04T23:59:59.000Z

    Advanced small modular reactors (aSMRs) can provide the United States with a safe, sustainable, and carbon-neutral energy source. The controllable day-to-day costs of aSMRs are expected to be dominated by operation and maintenance costs. Health and condition assessment coupled with online risk monitors can potentially enhance affordability of aSMRs through optimized operational planning and maintenance scheduling. Currently deployed risk monitors are an extension of probabilistic risk assessment (PRA). For complex engineered systems like nuclear power plants, PRA systematically combines event likelihoods and the probability of failure (POF) of key components, so that when combined with the magnitude of possible adverse consequences to determine risk. Traditional PRA uses population-based POF information to estimate the average plant risk over time. Currently, most nuclear power plants have a PRA that reflects the as-operated, as-modified plant; this model is updated periodically, typically once a year. Risk monitors expand on living PRA by incorporating changes in the day-by-day plant operation and configuration (e.g., changes in equipment availability, operating regime, environmental conditions). However, population-based POF (or population- and time-based POF) is still used to populate fault trees. Health monitoring techniques can be used to establish condition indicators and monitoring capabilities that indicate the component-specific POF at a desired point in time (or over a desired period), which can then be incorporated in the risk monitor to provide a more accurate estimate of the plant risk in different configurations. This is particularly important for active systems, structures, and components (SSCs) proposed for use in aSMR designs. These SSCs may differ significantly from those used in the operating fleet of light-water reactors (or even in LWR-based SMR designs). Additionally, the operating characteristics of aSMRs can present significantly different requirements, including the need to operate in different coolant environments, higher operating temperatures, and longer operating cycles between planned refueling and maintenance outages. These features, along with the relative lack of operating experience for some of the proposed advanced designs, may limit the ability to estimate event probability and component POF with a high degree of certainty. Incorporating real-time estimates of component POF may compensate for a relative lack of established knowledge about the long-term component behavior and improve operational and maintenance planning and optimization. The particular eccentricities of advanced reactors and small modular reactors provide unique challenges and needs for advanced instrumentation, control, and human-machine interface (ICHMI) techniques such as enhanced risk monitors (ERM) in aSMRs. Several features of aSMR designs increase the need for accurate characterization of the real-time risk during operation and maintenance activities. A number of technical gaps in realizing ERM exist, and these gaps are largely independent of the specific reactor technology. As a result, the development of a framework for ERM would enable greater situational awareness regardless of the specific class of reactor technology. A set of research tasks are identified in a preliminary research plan to enable the development, testing, and demonstration of such a framework. Although some aspects of aSMRs, such as specific operational characteristics, will vary and are not now completely defined, the proposed framework is expected to be relevant regardless of such uncertainty. The development of an ERM framework will provide one of the key technical developments necessary to ensure the economic viability of aSMRs.

  20. Reactivity control assembly for nuclear reactor

    DOE Patents [OSTI]

    Bollinger, Lawrence R. (Schenectady, NY)

    1984-01-01T23:59:59.000Z

    Reactivity control assembly for nuclear reactor comprises supports stacked above reactor core for holding control rods. Couplers associated with the supports and a vertically movable drive shaft have lugs at their lower ends for engagement with the supports.

  1. Depletion Analysis of Modular High Temperature Gas-cooled Reactor Loaded with LEU/Thorium Fuel

    SciTech Connect (OSTI)

    Sonat Sen; Gilles Youinou

    2013-02-01T23:59:59.000Z

    Thorium based fuel has been considered as an option to uranium-based fuel, based on considerations of resource utilization (Thorium is more widely available when compared to Uranium). The fertile isotope of Thorium (Th-232) can be converted to fissile isotope U-233 by neutron capture during the operation of a suitable nuclear reactor such as High Temperature Gas-cooled Reactor (HTGR). However, the fertile Thorium needs a fissile supporter to start and maintain the conversion process such as U-235 or Pu-239. This report presents the results of a study that analyzed the thorium utilization in a prismatic HTGR, namely Modular High Temperature Gas-Cooled Reactor (MHTGR) that was designed by General Atomics (GA). The collected for the modeling of this design come from Chapter 4 of MHTGR Preliminary Safety Information Document that GA sent to Department of Energy (DOE) on 1995. Both full core and unit cell models were used to perform this analysis using SCALE 6.1 and Serpent 1.1.18. Because of the long mean free paths (and migration lengths) of neutrons in HTRs, using a unit cell to represent a whole core can be non-trivial. The sizes of these cells were set to match the spectral index between unit cell and full core domains. It was found that for the purposes of this study an adjusted unit cell model is adequate. Discharge isotopics and one-group cross-sections were delivered to the transmutation analysis team. This report provides documentation for these calculations

  2. Reactor Subsystem Simulation for Nuclear Hybrid Energy Systems

    SciTech Connect (OSTI)

    Shannon Bragg-Sitton; J. Michael Doster; Alan Rominger

    2012-09-01T23:59:59.000Z

    Preliminary system models have been developed by Idaho National Laboratory researchers and are currently being enhanced to assess integrated system performance given multiple sources (e.g., nuclear + wind) and multiple applications (i.e., electricity + process heat). Initial efforts to integrate a Fortran-based simulation of a small modular reactor (SMR) with the balance of plant model have been completed in FY12. This initial effort takes advantage of an existing SMR model developed at North Carolina State University to provide initial integrated system simulation for a relatively low cost. The SMR subsystem simulation details are discussed in this report.

  3. Nuclear reactor multiphysics via bond graph formalism

    E-Print Network [OSTI]

    Sosnovsky, Eugeny

    2014-01-01T23:59:59.000Z

    This work proposes a simple and effective approach to modeling nuclear reactor multiphysics problems using bond graphs. Conventional multiphysics simulation paradigms normally use operator splitting, which treats the ...

  4. Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

    DOE Patents [OSTI]

    Schreiber, R.B.; Fero, A.H.; Sejvar, J.

    1997-12-16T23:59:59.000Z

    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 to form 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 valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor. 8 figs.

  5. Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

    DOE Patents [OSTI]

    Schreiber, Roger B. (Penn Twp., PA); Fero, Arnold H. (New Kensington, PA); Sejvar, James (Murrysville, PA)

    1997-01-01T23:59:59.000Z

    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 to form 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 valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor.

  6. Modular hybrid plasma reactor and related systems and methods

    DOE Patents [OSTI]

    Kong, Peter C.; Grandy, Jon D.; Detering, Brent A.

    2010-06-22T23:59:59.000Z

    A device, method and system for generating a plasma is disclosed wherein an electrical arc is established and the movement of the electrical arc is selectively controlled. In one example, modular units are coupled to one another to collectively define a chamber. Each modular unit may include an electrode and a cathode spaced apart and configured to generate an arc therebetween. A device, such as a magnetic or electromagnetic device, may be used to selectively control the movement of the arc about a longitudinal axis of the chamber. The arcs of individual modules may be individually controlled so as to exhibit similar or dissimilar motions about the longitudinal axis of the chamber. In another embodiment, an inlet structure may be used to selectively define the flow path of matter introduced into the chamber such that it travels in a substantially circular or helical path within the chamber.

  7. Small Modular Reactor: First of a Kind (FOAK) and Nth of a Kind (NOAK) Economic Analysis

    SciTech Connect (OSTI)

    Lauren M. Boldon; Piyush Sabharwall

    2014-08-01T23:59:59.000Z

    Small modular reactors (SMRs) refer to any reactor design in which the electricity generated is less than 300 MWe. Often medium sized reactors with power less than 700 MWe are also grouped into this category. Internationally, the development of a variety of designs for SMRs is booming with many designs approaching maturity and even in or nearing the licensing stage. It is for this reason that a generalized yet comprehensive economic model for first of a kind (FOAK) through nth of a kind (NOAK) SMRs based upon rated power, plant configuration, and the fiscal environment was developed. In the model, a particular project’s feasibility is assessed with regards to market conditions and by commonly utilized capital budgeting techniques, such as the net present value (NPV), internal rate of return (IRR), Payback, and more importantly, the levelized cost of energy (LCOE) for comparison to other energy production technologies. Finally, a sensitivity analysis was performed to determine the effects of changing debt, equity, interest rate, and conditions on the LCOE. The economic model is primarily applied to the near future water cooled SMR designs in the United States. Other gas cooled and liquid metal cooled SMR designs have been briefly outlined in terms of how the economic model would change. FOAK and NOAK SMR costs were determined for a site containing seven 180 MWe water cooled SMRs and compared to a site containing one 1260 MWe reactor. With an equal share of debt and equity and a 10% cost of debt and equity, the LCOE was determined to be $79 $84/MWh and $80/MWh for the SMR and large reactor sites, respectively. With a cost of equity of 15%, the SMR LCOE increased substantially to $103 $109/MWh. Finally, an increase in the equity share to 70% at the 15% cost of equity resulted in an even higher LCOE, demonstrating the large variation in results due to financial and market factors. The NPV and IRR both decreased with increasing LCOE. Unless the price of electricity increases along with the LCOE, the projects may become unprofitable. This is the case at the LCOE of $103 $109/MW, in which the NPV became negative. The IRR increased with increasing electricity price. Three cases, electric only base, storage—compressed air energy storage or pumped hydro, and hydrogen production, were performed incorporating SMRs into a nuclear wind natural gas hybrid energy system for the New York West Central region. The operational costs for three cases were calculated as $27/MWh, $25/MWh, and $28/MWh, respectively. A 3% increase in profits was demonstrated for the storage case over the electric only base case.

  8. Economic Analysis of the Modular Pebble Bed Reactor

    E-Print Network [OSTI]

    data found inexisting cost data found in ""Evaluation of the Gas Turbine HeliumEvaluation of the Gas Turbine Helium ReactorReactor"" -- DOEDOE--HTGRHTGR--9038090380 -- Dec. 1993 and compared against an the gas prices price in 1992 was assumed constant and did not increase. This study win 1992 was assumed

  9. Simulated nuclear reactor fuel assembly

    DOE Patents [OSTI]

    Berta, Victor T. (Idaho Falls, ID)

    1993-01-01T23:59:59.000Z

    An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

  10. Simulated nuclear reactor fuel assembly

    DOE Patents [OSTI]

    Berta, V.T.

    1993-04-06T23:59:59.000Z

    An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

  11. APPLICATION OF DATA ANALYSIS TECHNIQUES TO NUCLEAR REACTOR

    E-Print Network [OSTI]

    Kunz, Robert Francis

    1 APPLICATION OF DATA ANALYSIS TECHNIQUES TO NUCLEAR REACTOR SYSTEMS CODE ACCURACY ASSESSMENT) has been developed by the authors to provide quantitative comparisons between nuclear reactor systems. 1. INTRODUCTION In recent years, the commercial nuclear reactor industry has focused significant

  12. Integrating Safety, Operations, Security, and Safeguards (ISOSS) into the design of small modular reactors : a handbook.

    SciTech Connect (OSTI)

    Middleton, Bobby D.; Mendez, Carmen Margarita [Sociotecnia Solutions] [Sociotecnia Solutions

    2013-10-01T23:59:59.000Z

    The existing regulatory environment for nuclear reactors impacts both the facility design and the cost of operations once the facility is built. Delaying the consideration of regulatory requirements until late in the facility design - or worse, until after construction has begun - can result in costly retrofitting as well as increased operational costs to fulfill safety, security, safeguards, and emergency readiness requirements. Considering the scale and scope, as well as the latest design trends in the next generation of nuclear facilities, there is an opportunity to evaluate the regulatory requirements and optimize the design process for Small Modular Reactors (SMRs), as compared to current Light Water Reactors (LWRs). To this end, Sandia has embarked on an initiative to evaluate the interactions of regulations and operations as an approach to optimizing the design of SMR facilities, supporting operational efficiencies, as well as regulatory requirements. The early stages of this initiative consider two focus areas. The first focus area, reported by LaChance, et al. (2007), identifies the regulatory requirements established for the current fleet of LWR facilities regarding Safety, Security, Operations, Safeguards, and Emergency Planning, and evaluates the technical bases for these requirements. The second focus area, developed in this report, documents the foundations for an innovative approach that supports a design framework for SMR facilities that incorporates the regulatory environment, as well as the continued operation of the facility, into the early design stages, eliminating the need for costly retrofitting and additional operating personnel to fulfill regulatory requirements. The work considers a technique known as Integrated Safety, Operations, Security and Safeguards (ISOSS) (Darby, et al., 2007). In coordination with the best practices of industrial operations, the goal of this effort is to develop a design framework that outlines how ISOSS requirements can be incorporated into the pre-conceptual through early facility design stages, seeking a cost-effective design that meets both operational efficiencies and the regulatory environment. The larger scope of the project, i.e., in future stages, includes the identification of potentially conflicting requirements identified by the ISOSS framework, including an analysis of how regulatory requirements may be changed to account for the intrinsic features of SMRs.

  13. Studies on the closed-loop digital control of multi-modular reactors

    SciTech Connect (OSTI)

    Bernard, J.A. (Massachusetts Inst. of Tech., Cambridge, MA (United States). Nuclear Reactor Lab.); Henry, A.F.; Lanning, D.D.; Meyer, J.E. (Massachusetts Inst. of Tech., Cambridge, MA (United States). Dept. of Nuclear Engineering)

    1992-11-01T23:59:59.000Z

    This report describes the theoretical development and the evaluation via both experiment and simulation of digital methods for the closed-loop control of power, temperature, and steam generator level in multi-modular reactors. The major conclusion of the research reported here is that the technology is currently available to automate many aspects of the operation of multi-modular plants. This will in turn minimize the number of required personnel and thus contain both operating and personnel costs, allow each module to be operated at a different power level thereby staggering the times at which refuelings would be needed, and maintain the competitiveness of US industry relative to foreign vendors who are developing and applying advanced control concepts. The technology described in this report is appropriate to the proposed multi-modular reactor designs and to present-generation pressurized water reactors. Its extension to boiling water reactors is possible provided that the commitment is made to create a real-time model of a BWR. The work reported here was performed by the Massachusetts Institute of Technology (MIT) under contract to the Oak Ridge National Laboratory (ORNL) and to the United States Department of Energy (Division of Industry and University Programs, Contract No. DE-FG07-90ER12930.)

  14. Studies on the closed-loop digital control of multi-modular reactors. Final report

    SciTech Connect (OSTI)

    Bernard, J.A. [Massachusetts Inst. of Tech., Cambridge, MA (United States). Nuclear Reactor Lab.; Henry, A.F.; Lanning, D.D.; Meyer, J.E. [Massachusetts Inst. of Tech., Cambridge, MA (United States). Dept. of Nuclear Engineering

    1992-11-01T23:59:59.000Z

    This report describes the theoretical development and the evaluation via both experiment and simulation of digital methods for the closed-loop control of power, temperature, and steam generator level in multi-modular reactors. The major conclusion of the research reported here is that the technology is currently available to automate many aspects of the operation of multi-modular plants. This will in turn minimize the number of required personnel and thus contain both operating and personnel costs, allow each module to be operated at a different power level thereby staggering the times at which refuelings would be needed, and maintain the competitiveness of US industry relative to foreign vendors who are developing and applying advanced control concepts. The technology described in this report is appropriate to the proposed multi-modular reactor designs and to present-generation pressurized water reactors. Its extension to boiling water reactors is possible provided that the commitment is made to create a real-time model of a BWR. The work reported here was performed by the Massachusetts Institute of Technology (MIT) under contract to the Oak Ridge National Laboratory (ORNL) and to the United States Department of Energy (Division of Industry and University Programs, Contract No. DE-FG07-90ER12930.)

  15. Proliferation Resistant Nuclear Reactor Fuel

    SciTech Connect (OSTI)

    Gray, L W; Moody, K J; Bradley, K S; Lorenzana, H E

    2011-02-18T23:59:59.000Z

    Global appetite for fission power is projected to grow dramatically this century, and for good reason. Despite considerable research to identify new sources of energy, fission remains the most plentiful and practical alternative to fossil fuels. The environmental challenges of fossil fuel have made the fission power option increasingly attractive, particularly as we are forced to rely on reserves in ecologically fragile or politically unstable corners of the globe. Caught between a globally eroding fossil fuel reserve as well as the uncertainty and considerable costs in the development of fusion power, most of the world will most likely come to rely on fission power for at least the remainder of the 21st century. Despite inevitable growth, fission power faces enduring challenges in sustainability and security. One of fission power's greatest hurdles to universal acceptance is the risk of potential misuse for nefarious purposes of fissionable byproducts in spent fuel, such as plutonium. With this issue in mind, we have discussed intrinsic concepts in this report that are motivated by the premise that the utility, desirability, and applicability of nuclear materials can be reduced. In a general sense, the intrinsic solutions aim to reduce or eliminate the quantity of existing weapons usable material; avoid production of new weapons-usable material through enrichment, breeding, extraction; or employ engineering solutions to make the fuel cycle less useful or more difficult for producing weapons-usable material. By their nature, these schemes require modifications to existing fuel cycles. As such, the concomitants of these modifications require engagement from the nuclear reactor and fuel-design community to fully assess their effects. Unfortunately, active pursuit of any scheme that could further complicate the spread of domestic nuclear power will probably be understandably unpopular. Nevertheless, the nonproliferation and counterterrorism issues are paramount, and we posit that the exploration, development, and implementation of intrinsic mechanisms such as discussed here are part of a balanced approach aimed at preventing the misuse of nuclear material for nuclear-energy applications.

  16. Physics-based multiscale coupling for full core nuclear reactor...

    Office of Scientific and Technical Information (OSTI)

    multiscale coupling for full core nuclear reactor simulation Numerical simulation of nuclear reactors is a key technology in the quest for improvements in efficiency, safety,...

  17. SciTech Connect: Fundamental aspects of nuclear reactor fuel...

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

    Fundamental aspects of nuclear reactor fuel elements Citation Details In-Document Search Title: Fundamental aspects of nuclear reactor fuel elements You are accessing a document...

  18. Digital computer operation of a nuclear reactor

    DOE Patents [OSTI]

    Colley, Robert W. (Richland, WA)

    1984-01-01T23:59:59.000Z

    A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

  19. Digital computer operation of a nuclear reactor

    DOE Patents [OSTI]

    Colley, R.W.

    1982-06-29T23:59:59.000Z

    A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

  20. Liquid metal cooled nuclear reactor plant system

    DOE Patents [OSTI]

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

    1993-01-01T23:59:59.000Z

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

  1. ME 361E Nuclear Reactor Engineering ABET EC2000 syllabus

    E-Print Network [OSTI]

    Ben-Yakar, Adela

    ME 361E ­ Nuclear Reactor Engineering Page 1 ABET EC2000 syllabus ME 361E ­ Nuclear Reactor students should be able to: · Compare and contrast numerous nuclear reactor designs · Calculate the effects of nuclear fuel burnup · Summarize the mechanism that affect the control of a nuclear reactor Topics Covered

  2. Nuclear reactor control apparatus. [FBR

    SciTech Connect (OSTI)

    Sridhar, B.N.

    1981-04-16T23:59:59.000Z

    Nuclear reactor safety rod release apparatus comprises a ring which carries detents normally positioned in an annular recess in outer side of the rod, the ring being held against the lower end of a drive shaft by magnetic force exerted by a solenoid carried by the drive shaft. When the solenoid is de-energized, the detent-carrying ring drops until the detents contact a cam surface associated with the lower end of the drive shaft, at which point the detents are cammed out of the recess in the safety rod to release the rod from the drive shaft. In preferred embodiments of the invention, an additional latch is provided to release a lower portion of a safety rod under conditions that may interfere with movement of the entire rod.

  3. Nuclear reactor internals alignment configuration

    DOE Patents [OSTI]

    Gilmore, Charles B. (Greensburg, PA); Singleton, Norman R. (Murrysville, PA)

    2009-11-10T23:59:59.000Z

    An alignment system that employs jacking block assemblies and alignment posts around the periphery of the top plate of a nuclear reactor lower internals core shroud to align an upper core plate with the lower internals and the core shroud with the core barrel. The distal ends of the alignment posts are chamfered and are closely received within notches machined in the upper core plate at spaced locations around the outer circumference of the upper core plate. The jacking block assemblies are used to center the core shroud in the core barrel and the alignment posts assure the proper orientation of the upper core plate. The alignment posts may alternately be formed in the upper core plate and the notches may be formed in top plate.

  4. Nuclear reactor composite fuel assembly

    DOE Patents [OSTI]

    Burgess, Donn M. (Richland, WA); Marr, Duane R. (West Richland, WA); Cappiello, Michael W. (Richland, WA); Omberg, Ronald P. (Richland, WA)

    1980-01-01T23:59:59.000Z

    A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

  5. Reference modular High Temperature Gas-Cooled Reactor Plant: Concept description report

    SciTech Connect (OSTI)

    Not Available

    1986-10-01T23:59:59.000Z

    This report provides a summary description of the Modular High Temperature Gas-Cooled Reactor (MHTGR) concept and interim results of assessments of costs, safety, constructibility, operability, maintainability, and availability. Conceptual design of this concept was initiated in October 1985 and is scheduled for completion in 1987. Participating industrial contractors are Bechtel National, Inc. (BNI), Stone and Webster Engineering Corporation (SWEC), GA Technologies, Inc. (GA), General Electric Co. (GE), and Combustion Engineering, Inc. (C-E).

  6. Systems Issues in Nuclear Reactor Safety

    E-Print Network [OSTI]

    de Weck, Olivier L.

    Systems Issues in Nuclear Reactor Safety Commissioner George ApostolakisCommissioner George Apostolakis U.S. Nuclear Regulatory Commission CmrApostolakis@nrc.gov MIT SDM Conference on Systems Thinking, source, and special nuclear materials to ensure adequate protection of public health and safety, 3

  7. Fast-acting nuclear reactor control device

    DOE Patents [OSTI]

    Kotlyar, Oleg M. (Idaho Falls, ID); West, Phillip B. (Idaho Falls, ID)

    1993-01-01T23:59:59.000Z

    A fast-acting nuclear reactor control device for moving and positioning a fety control rod to desired positions within the core of the reactor between a run position in which the safety control rod is outside the reactor core, and a shutdown position in which the rod is fully inserted in the reactor core. The device employs a hydraulic pump/motor, an electric gear motor, and solenoid valve to drive the safety control rod into the reactor core through the entire stroke of the safety control rod. An overrunning clutch allows the safety control rod to freely travel toward a safe position in the event of a partial drive system failure.

  8. Shutdown system for a nuclear reactor

    DOE Patents [OSTI]

    Groh, E.F.; Olson, A.P.; Wade, D.C.; Robinson, B.W.

    1984-06-05T23:59:59.000Z

    An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion. 8 figs.

  9. Shutdown system for a nuclear reactor

    DOE Patents [OSTI]

    Groh, Edward F. (Naperville, IL); Olson, Arne P. (Western Springs, IL); Wade, David C. (Naperville, IL); Robinson, Bryan W. (Oak Lawn, IL)

    1984-01-01T23:59:59.000Z

    An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion.

  10. Nuclear reactor safeguards and monitoring with antineutrino detectors A. Bernsteina)

    E-Print Network [OSTI]

    Gratta, Giorgio

    Nuclear reactor safeguards and monitoring with antineutrino detectors A. Bernsteina) Sandia of nuclear reactor types, including power reactors, research reactors, and plutonium production reactors-understood principles that govern the core's evolution in time, can be used to determine whether the reactor is being

  11. Gas-cooled nuclear reactor

    DOE Patents [OSTI]

    Peinado, Charles O. (La Jolla, CA); Koutz, Stanley L. (San Diego, CA)

    1985-01-01T23:59:59.000Z

    A gas-cooled nuclear reactor includes a central core located in the lower portion of a prestressed concrete reactor vessel. Primary coolant gas flows upward through the core and into four overlying heat-exchangers wherein stream is generated. During normal operation, the return flow of coolant is between the core and the vessel sidewall to a pair of motor-driven circulators located at about the bottom of the concrete pressure vessel. The circulators repressurize the gas coolant and return it back to the core through passageways in the underlying core structure. If during emergency conditions the primary circulators are no longer functioning, the decay heat is effectively removed from the core by means of natural convection circulation. The hot gas rising through the core exits the top of the shroud of the heat-exchangers and flows radially outward to the sidewall of the concrete pressure vessel. A metal liner covers the entire inside concrete surfaces of the concrete pressure vessel, and cooling tubes are welded to the exterior or concrete side of the metal liner. The gas coolant is in direct contact with the interior surface of the metal liner and transfers its heat through the metal liner to the liquid coolant flowing through the cooling tubes. The cooler gas is more dense and creates a downward convection flow in the region between the core and the sidewall until it reaches the bottom of the concrete pressure vessel when it flows radially inward and up into the core for another pass. Water is forced to flow through the cooling tubes to absorb heat from the core at a sufficient rate to remove enough of the decay heat created in the core to prevent overheating of the core or the vessel.

  12. Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who

    SciTech Connect (OSTI)

    Forsberg, C.W.; Reich, W.J.

    1991-09-01T23:59:59.000Z

    The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactor concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs.

  13. Today and Future Neutrino Experiments at Krasnoyarsk Nuclear Reactor

    E-Print Network [OSTI]

    Yu. V. Kozlov; S. V. Khalturtsev; I. N. Machulin; A. V. Martemyanov; V. P. Martemyanov; A. A. Sabelnikov; V. G. Tarasenkov; E. V. Turbin; V. N. Vyrodov; L. A. Popeko; A. V. Cherny; G. A. Shishkina

    1999-12-22T23:59:59.000Z

    The results of undergoing experiments and new experiment propositions at Krasnoyarsk underground nuclear reactor are presented

  14. Nuclear facilities: criteria for the design and operation of ventilation systems for nuclear installations other than nuclear reactors

    E-Print Network [OSTI]

    International Organization for Standardization. Geneva

    2004-01-01T23:59:59.000Z

    Nuclear facilities: criteria for the design and operation of ventilation systems for nuclear installations other than nuclear reactors

  15. Perspective unconventional means for nuclear reactor control

    SciTech Connect (OSTI)

    Ionaitis, R.R.

    1993-12-31T23:59:59.000Z

    The concept of reliable shutdown of nuclear reactors demands application of engineering control means on the basis of principles, safety, safe failure, redundancy, independence, variety, defence in depth, and intrinsical safety. This report describes application of control methods.

  16. 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

    2003-06-16T23:59:59.000Z

    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

  17. Uncertainties in the Anti-neutrino Production at Nuclear Reactors

    E-Print Network [OSTI]

    Djurcic, Zelimir

    2009-01-01T23:59:59.000Z

    reactors are determined from thermal power measure- ments and ?ssion rate calculations.of a reactor’s ther- mal power is given by a calculation ofCALCULATIONS During the power cycle of a nuclear reactor,

  18. Large Scale Weather Control Using Nuclear Reactors

    E-Print Network [OSTI]

    Singh-Modgil, M

    2002-01-01T23:59:59.000Z

    It is pointed out that controlled release of thermal energy from fission type nuclear reactors can be used to alter weather patterns over significantly large geographical regions. (1) Nuclear heat creates a low pressure region, which can be used to draw moist air from oceans, onto deserts. (2) Creation of low pressure zones over oceans using Nuclear heat can lead to Controlled Cyclone Creation (CCC).(3) Nuclear heat can also be used to melt glaciers and control water flow in rivers.

  19. Large Scale Weather Control Using Nuclear Reactors

    E-Print Network [OSTI]

    Moninder Singh Modgil

    2002-10-02T23:59:59.000Z

    It is pointed out that controlled release of thermal energy from fission type nuclear reactors can be used to alter weather patterns over significantly large geographical regions. (1) Nuclear heat creates a low pressure region, which can be used to draw moist air from oceans, onto deserts. (2) Creation of low pressure zones over oceans using Nuclear heat can lead to Controlled Cyclone Creation (CCC).(3) Nuclear heat can also be used to melt glaciers and control water flow in rivers.

  20. Westinghouse Small Modular Reactor passive safety system response to postulated events

    SciTech Connect (OSTI)

    Smith, M. C.; Wright, R. F. [Westinghouse Electric Company, 600 Cranberry Woods Drive (United States)

    2012-07-01T23:59:59.000Z

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor. This paper is part of a series of four describing the design and safety features of the Westinghouse SMR. This paper focuses in particular upon the passive safety features and the safety system response of the Westinghouse SMR. The Westinghouse SMR design incorporates many features to minimize the effects of, and in some cases eliminates the possibility of postulated accidents. The small size of the reactor and the low power density limits the potential consequences of an accident relative to a large plant. The integral design eliminates large loop piping, which significantly reduces the flow area of postulated loss of coolant accidents (LOCAs). The Westinghouse SMR containment is a high-pressure, compact design that normally operates at a partial vacuum. This facilitates heat removal from the containment during LOCA events. The containment is submerged in water which also aides the heat removal and provides an additional radionuclide filter. The Westinghouse SMR safety system design is passive, is based largely on the passive safety systems used in the AP1000{sup R} reactor, and provides mitigation of all design basis accidents without the need for AC electrical power for a period of seven days. Frequent faults, such as reactivity insertion events and loss of power events, are protected by first shutting down the nuclear reaction by inserting control rods, then providing cold, borated water through a passive, buoyancy-driven flow. Decay heat removal is provided using a layered approach that includes the passive removal of heat by the steam drum and independent passive heat removal system that transfers heat from the primary system to the environment. Less frequent faults such as loss of coolant accidents are mitigated by passive injection of a large quantity of water that is readily available inside containment. An automatic depressurization system is used to reduce the reactor pressure in a controlled manner to facilitate the passive injection. Long-term decay heat removal is accomplished using the passive heat removal systems augmented by heat transfer through the containment vessel to the environment. The passive injection systems are designed so that the fuel remains covered and effectively cooled throughout the event. Like during the frequent faults, the passive systems provide effective cooling without the need for ac power for seven days following the accident. Connections are available to add additional water to indefinitely cool the plant. The response of the safety systems of the Westinghouse SMR to various initiating faults has been examined. Among them, two accidents; an extended station blackout event, and a LOCA event have been evaluated to demonstrate how the plant will remain safe in the unlikely event that either should occur. (authors)

  1. Modular Hybrid Plasma Reactor for Low Cost Bulk Production of Nanomaterials

    SciTech Connect (OSTI)

    Peter C. Kong

    2011-12-01T23:59:59.000Z

    INL developed a bench scale modular hybrid plasma system for gas phase nanomaterials synthesis. The system was being optimized for WO3 nanoparticles production and scale model projection to a 300 kW pilot system. During the course of technology development many modifications had been done to the system to resolve technical issues that had surfaced and also to improve the performance. All project tasks had been completed except 2 optimization subtasks. These 2 subtasks, a 4-hour and an 8-hour continuous powder production runs at 1 lb/hr powder feeding rate, were unable to complete due to technical issues developed with the reactor system. The 4-hour run had been attempted twice and both times the run was terminated prematurely. The modular electrode for the plasma system was significantly redesigned to address the technical issues. Fabrication of the redesigned modular electrodes and additional components had been completed at the end of the project life. However, not enough resource was available to perform tests to evaluate the performance of the new modifications. More development work would be needed to resolve these problems prior to scaling. The technology demonstrated a surprising capability of synthesizing a single phase of meta-stable delta-Al2O3 from pure alpha-phase large Al2O3 powder. The formation of delta-Al2O3 was surprising because this phase is meta-stable and only formed between 973-1073 K, and delta-Al2O3 is very difficult to synthesize as a single phase. Besides the specific temperature window to form this phase, this meta-stable phase may have been stabilized by nanoparticle size formed in a high temperature plasma process. This technology may possess the capability to produce unusual meta-stable nanophase materials that would be otherwise difficult to produce by conventional methods. A 300 kW INL modular hybrid plasma pilot scale model reactor had been projected using the experimental data from PPG Industries 300 kW hot wall plasma reactor. The projected size of the INL 300 kW pilot model reactor would be about 15% that of the PPG 300 kW hot wall plasma reactor. Including the safety net factor the projected INL pilot reactor size would be 25-30% of the PPG 300 kW hot wall plasma pilot reactor. Due to the modularity of the INL plasma reactor and the energy cascading effect from the upstream plasma to the downstream plasma the energy utilization is more efficient in material processing. It is envisioning that the material through put range for the INL pilot reactor would be comparable to the PPG 300 kW pilot reactor but the energy consumption would be lower. The INL hybrid plasma technology is rather close to being optimized for scaling to a pilot system. More near term development work is still needed to complete the process optimization before pilot scaling.

  2. Nuclear reactor vessel fuel thermal insulating barrier

    DOE Patents [OSTI]

    Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

    2013-03-19T23:59:59.000Z

    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.

  3. Fast-acting nuclear reactor control device

    SciTech Connect (OSTI)

    Kotlyar, O.M.; West, P.B.

    1993-08-03T23:59:59.000Z

    A fast-acting nuclear reactor control device is described for controlling a safety control rod within the core of a nuclear reactor, the reactor controlled by a reactor control system, the device comprising: a safety control rod drive shaft and an electromagnetic clutch co-axial with the drive shaft operatively connected to the safety control rod for driving and positioning the safety control rod within or without the reactor core during reactor operation, the safety rod being oriented in a substantially vertical position to allow the rod to fall into the reactor core under the influence of gravity during shutdown of the reactor; the safety control rod drive shaft further operatively connected to a hydraulic pump such that operation of the drive shaft simultaneously drives and positions the safety control rod and operates the hydraulic pump such that a hydraulic fluid is forced into an accumulator, filling the accumulator with oil for the storage and supply of primary potential energy for safety control rod insertion such that the release of potential energy in the accumulator causes hydraulic fluid to flow through the hydraulic pump, converting the hydraulic pump to a hydraulic motor having speed and power capable of full length insertion and high speed driving of the safety control rod into the reactor core; a solenoid valve interposed between the hydraulic pump and the accumulator, said solenoid valve being a normally open valve, actuated to close when the safety control rod is out of the reactor during reactor operation; and further wherein said solenoid opens in response to a signal from the reactor control system calling for shutdown of the reactor and rapid insertion of the safety control rod into the reactor core, such that the opening of the solenoid releases the potential energy in the accumulator to place the safety control rod in a safe shutdown position.

  4. Probabilistic accident analysis of the Pebble Bed modular Reactor for use with risk informed regulation

    E-Print Network [OSTI]

    Savkina, Marina D., 1973-

    2004-01-01T23:59:59.000Z

    One of the major challenges to the successful deployment of new nuclear plants in the United States is the regulatory process, which is largely based on water-reactor technology. While ongoing and expected efforts to license ...

  5. Nuclear reactor construction with bottom supported reactor vessel

    DOE Patents [OSTI]

    Sharbaugh, John E. (Bullskin Township, Fayette County, PA)

    1987-01-01T23:59:59.000Z

    An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment structure base mat so as to insulate the reactor vessel bottom end wall from the containment structure base mat and allow the reactor vessel bottom end wall to freely expand as it heats up while providing continuous support thereof. Further, a deck is supported upon the side wall of the containment structure above the top open end of the reactor vessel, and a plurality of serially connected extendible and retractable annular bellows extend between the deck and the top open end of the reactor vessel and flexibly and sealably interconnect the reactor vessel at its top end to the deck. An annular guide ring is disposed on the containment structure and extends between its side wall and the top open end of the reactor vessel for providing lateral support of the reactor vessel top open end by limiting imposition of lateral loads on the annular bellows by the occurrence of a lateral seismic event.

  6. Heat dissipating nuclear reactor with metal liner

    DOE Patents [OSTI]

    Gluekler, E.L.; Hunsbedt, A.; Lazarus, J.D.

    1985-11-21T23:59:59.000Z

    A nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel is described in this disclosure. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

  7. Heat dissipating nuclear reactor with metal liner

    DOE Patents [OSTI]

    Gluekler, Emil L. (San Jose, CA); Hunsbedt, Anstein (Los Gatos, CA); Lazarus, Jonathan D. (Sunnyvale, CA)

    1987-01-01T23:59:59.000Z

    Disclosed is a nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

  8. Reactivity control assembly for nuclear reactor. [LMFBR

    DOE Patents [OSTI]

    Bollinger, L.R.

    1982-03-17T23:59:59.000Z

    This invention, which resulted from a contact with the United States Department of Energy, relates to a control mechanism for a nuclear reactor and, more particularly, to an assembly for selectively shifting different numbers of reactivity modifying rods into and out of the core of a nuclear reactor. It has been proposed heretofore to control the reactivity of a breeder reactor by varying the depth of insertion of control rods (e.g., rods containing a fertile material such as ThO/sub 2/) in the core of the reactor, thereby varying the amount of neutron-thermalizing coolant and the amount of neutron-capturing material in the core. This invention relates to a mechanism which can advantageously be used in this type of reactor control system.

  9. UPDATE ON SMALL MODULAR REACTORS DYNAMIC SYSTEM MODELING TOOL Molten Salt Cooled Architecture

    SciTech Connect (OSTI)

    Hale, Richard Edward [ORNL; Cetiner, Sacit M [ORNL; Fugate, David L [ORNL; Qualls, A L [ORNL; Borum, Robert C [ORNL; Chaleff, Ethan S [ORNL; Rogerson, Doug W [ORNL; Batteh, John J [Modelon Corporation; Tiller, Michael M. [Xogeny Corporation

    2014-08-01T23:59:59.000Z

    The Small Modular Reactor (SMR) Dynamic System Modeling Tool project is in the third year of development. The project is designed to support collaborative modeling and study of various advanced SMR (non-light water cooled) concepts, including the use of multiple coupled reactors at a single site. The objective of the project is to provide a common simulation environment and baseline modeling resources to facilitate rapid development of dynamic advanced reactor SMR models, ensure consistency among research products within the Instrumentation, Controls, and Human-Machine Interface (ICHMI) technical area, and leverage cross-cutting capabilities while minimizing duplication of effort. The combined simulation environment and suite of models are identified as the Modular Dynamic SIMulation (MoDSIM) tool. The critical elements of this effort include (1) defining a standardized, common simulation environment that can be applied throughout the program, (2) developing a library of baseline component modules that can be assembled into full plant models using existing geometry and thermal-hydraulic data, (3) defining modeling conventions for interconnecting component models, and (4) establishing user interfaces and support tools to facilitate simulation development (i.e., configuration and parameterization), execution, and results display and capture.

  10. Potential Application of Electrical Signature Analysis Methods for Monitoring Small Modular Reactor Components

    SciTech Connect (OSTI)

    Damiano, Brian [ORNL] [ORNL; Tucker Jr, Raymond W [ORNL] [ORNL; Haynes, Howard D [ORNL] [ORNL

    2010-01-01T23:59:59.000Z

    This paper will describe the technical basis behind ESA and why we consider it a viable SMR condition monitoring technology. Concepts are presented of how ESA could be applied to monitor two candidate small modular reactor components: the main coolant pumps and the control rod drives. We believe the general health of these two components can be monitored and trended over time, using ESA methods. Our optimism is based on over two decades of ESA development and testing on a wide variety of components and systems, many of which have similar operational features to the main coolant pumps and control rod drives.

  11. MHTGR (modular high-temperature gas-cooled reactor) control: A non-safety related system

    SciTech Connect (OSTI)

    Rodriguez, C.; Swart, F.

    1988-06-01T23:59:59.000Z

    The modular high-temperature gas-cooled reactor (MHTGR) design meets stringent top-level safety regulatory criteria and user requirements that call for high plant availability and no disruption of the public's day to day activities during normal and off-normal operation of the plant. These requirements lead to a plant design that relies mainly on physical properties and passive design features to ensure plant safety regardless of operator actions, plus simplicity and automation to ensure high plant availability and lower cost of operations. The plant does not require safety-related operator actions, and it does not require the control room to be safety related.

  12. Optimally moderated nuclear fission reactor and fuel source therefor

    DOE Patents [OSTI]

    Ougouag, Abderrafi M. (Idaho Falls, ID); Terry, William K. (Shelley, ID); Gougar, Hans D. (Idaho Falls, ID)

    2008-07-22T23:59:59.000Z

    An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

  13. White paper report on using nuclear reactors to search for a value of theta13

    E-Print Network [OSTI]

    2004-01-01T23:59:59.000Z

    PAPER REPORT on Using Nuclear Reactors to Search for a valuetimely new experiment at a nuclear reactor sensitive to theand judicious choice of a nuclear reactor. The dominant

  14. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01T23:59:59.000Z

    L. J. Hamilton Nuclear Reactor Analysis John Wiley and Sons,R. J. Neuhold, Introductury Nuclear Reactor Dynamics. ANSL. J. Hamilton Nuclear Reactor Analysis John Wiley and Sons,

  15. Technical Readiness and Gaps Analysis of Commercial Optical Materials and Measurement Systems for Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Anheier, Norman C.; Suter, Jonathan D.; Qiao, Hong (Amy); Andersen, Eric S.; Berglin, Eric J.; Bliss, Mary; Cannon, Bret D.; Devanathan, Ramaswami; Mendoza, Albert; Sheen, David M.

    2013-08-06T23:59:59.000Z

    This report intends to support Department of Energy’s Office of Nuclear Energy (DOE-NE) Nuclear Energy Research and Development Roadmap and industry stakeholders by evaluating optical-based instrumentation and control (I&C) concepts for advanced small modular reactor (AdvSMR) applications. These advanced designs will require innovative thinking in terms of engineering approaches, materials integration, and I&C concepts to realize their eventual viability and deployability. The primary goals of this report include: 1. Establish preliminary I&C needs, performance requirements, and possible gaps for AdvSMR designs based on best available published design data. 2. Document commercial off-the-shelf (COTS) optical sensors, components, and materials in terms of their technical readiness to support essential AdvSMR in-vessel I&C systems. 3. Identify technology gaps by comparing the in-vessel monitoring requirements and environmental constraints to COTS optical sensor and materials performance specifications. 4. Outline a future research, development, and demonstration (RD&D) program plan that addresses these gaps and develops optical-based I&C systems that enhance the viability of future AdvSMR designs. The development of clean, affordable, safe, and proliferation-resistant nuclear power is a key goal that is documented in the Nuclear Energy Research and Development Roadmap. This roadmap outlines RD&D activities intended to overcome technical, economic, and other barriers, which currently limit advances in nuclear energy. These activities will ensure that nuclear energy remains a viable component to this nation’s energy security.

  16. Fast-acting nuclear reactor control device

    SciTech Connect (OSTI)

    Kotlyar, O.M.; West, P.B.

    1992-12-31T23:59:59.000Z

    This invention consists of a fast-acting nuclear reactor control device for moving and positioning a safety control rod to desired elevations within the core of the reactor between a run position in which the safety control rod is outside the reactor core, and a shutdown position in which the rod is fully inserted in the reactor core. The device employs a hydraulic pump motor, an electric gear motor, and a solenoid valve to drive the safety control rod into the reactor core through the entire stroke of the safety control rod. An overrunning clutch, allows the safety control rod to freely travel toward a safe position in the event of a partial drive system failure.

  17. Numerical Study on Crossflow Printed Circuit Heat Exchanger for Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Yoon, Su-Jong [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Sabharwall, Piyush [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Kim, Eung-Soo [Seoul National Univ., Seoul (Korea, Republic of)

    2014-03-01T23:59:59.000Z

    Various fluids such as water, gases (helium), molten salts (FLiNaK, FLiBe) and liquid metal (sodium) are used as a coolant of advanced small modular reactors (SMRs). The printed circuit heat exchanger (PCHE) has been adopted as the intermediate and/or secondary heat exchanger of SMR systems because this heat exchanger is compact and effective. The size and cost of PCHE can be changed by the coolant type of each SMR. In this study, the crossflow PCHE analysis code for advanced small modular reactor has been developed for the thermal design and cost estimation of the heat exchanger. The analytical solution of single pass, both unmixed fluids crossflow heat exchanger model was employed to calculate a two dimensional temperature profile of a crossflow PCHE. The analytical solution of crossflow heat exchanger was simply implemented by using built in function of the MATLAB program. The effect of fluid property uncertainty on the calculation results was evaluated. In addition, the effect of heat transfer correlations on the calculated temperature profile was analyzed by taking into account possible combinations of primary and secondary coolants in the SMR systems. Size and cost of heat exchanger were evaluated for the given temperature requirement of each SMR.

  18. Nuclear reactor fissile isotopes antineutrino spectra

    E-Print Network [OSTI]

    V. Sinev

    2012-07-30T23:59:59.000Z

    Positron spectrum from inverse beta decay reaction on proton was measured in 1988-1990 as a result of neutrino exploration experiment. The measured spectrum has the largest statistics and lowest energy threshold between other neutrino experiments made that time at nuclear reactors. On base of the positron spectrum the standard antineutrino spectrum for typical reactor fuel composition was restored. In presented analysis the partial spectra forming this standard spectrum were extracted using specific method. They could be used for neutrino experiments data analysis made at any fuel composition of reactor core.

  19. Theta 13 Determination with Nuclear Reactors

    E-Print Network [OSTI]

    F. Dalnoki-Veress

    2004-06-24T23:59:59.000Z

    Recently there has been a lot of interest around the world in the use of nuclear reactors to measure theta 13, the last undetermined angle in the 3-neutrino mixing scenario. In this paper the motivations for theta 13 measurement using short baseline nuclear reactor experiments are discussed. The features of such an experiment are described in the context of Double Chooz, which is a new project planned to start data-taking in 2008, and to reach a sensitivity of sinsq(2 theta 13) < 0.03.

  20. 22.312 Engineering of Nuclear Reactors, Fall 2002

    E-Print Network [OSTI]

    Todreas, Neil E.

    Engineering principles of nuclear reactors, emphasizing power reactors. Power plant thermodynamics, reactor heat generation and removal (single-phase as well as two-phase coolant flow and heat transfer), and structural ...

  1. 22.312 Engineering of Nuclear Reactors, Fall 2004

    E-Print Network [OSTI]

    Buongiorno, Jacopo, 1971-

    Engineering principles of nuclear reactors, emphasizing power reactors. Power plant thermodynamics, reactor heat generation and removal (single-phase as well as two-phase coolant flow and heat transfer), and structural ...

  2. NUCLEAR POWER AND RESEARCH REACTORS 1939 1942 1943 1944

    E-Print Network [OSTI]

    Pennycook, Steve

    #12;#12;11 #12;2 NUCLEAR POWER AND RESEARCH REACTORS 1939 1942 1943 1944 Nuclear fission discovered Oak Ridge selected as site for World War II Manhattan Project First sustained and controlled nuclear 430 nuclear power reactors are operating in the world, and 103 nuclear power plants produce 20

  3. Nuclear Regulatory Commission Handling of Beyond Design Basis Events for Nuclear Power Reactors

    Broader source: Energy.gov [DOE]

    Presenter: Bill Reckley, Chief, Policy and Support Branch, Japan Lessons-Learned Project Directorate, Office of Nuclear Reactor Regulation, US Nuclear Regulatory Commission US Nuclear Regulatory Commission

  4. Current Abstracts Nuclear Reactors and Technology

    SciTech Connect (OSTI)

    Bales, J.D.; Hicks, S.C. [eds.

    1993-01-01T23:59:59.000Z

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  5. Nuclear reactor shutdown control rod assembly

    DOE Patents [OSTI]

    Bilibin, Konstantin (North Hollywood, CA)

    1988-01-01T23:59:59.000Z

    A temperature responsive, self-actuated nuclear reactor shutdown control rod assembly 10. The upper end 18 of a lower drive line 17 fits within the lower end of an upper drive line 12. The lower end (not shown) of the lower drive line 17 is connected to a neutron absorber. During normal temperature conditions the lower drive line 17 is supported by detent means 22,26. When an overtemperature condition occurs thermal actuation means 34 urges ring 26 upwardly sufficiently to allow balls 22 to move radially outwardly thereby allowing lower drive line 17 to move downwardly toward the core of the nuclear reactor resulting in automatic reduction of the reactor powder.

  6. Three Investment Scenarios for Future Nuclear Reactors in Europe

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    1 Three Investment Scenarios for Future Nuclear Reactors in Europe Bianka SHOAI TEHRANI CEA nuclear reactors within a few decades (2040), several events and drivers could question this possibility or detrimental to future nuclear reactors compared with other technologies and according to four main investment

  7. ARTIGO INTERNET Professores visitam o maior reactor de Fuso Nuclear

    E-Print Network [OSTI]

    Instituto de Sistemas e Robotica

    ARTIGO INTERNET Professores visitam o maior reactor de Fusão Nuclear in http reactor de Fusão Nuclear Experiência aproxima investigação das futuras gerações Doze professores do ensino secundário visitaram o maior reactor de fusão nuclear da Terra (JET), no Reino Unido, na semana passada

  8. MA50177: Scientific Computing Nuclear Reactor Simulation Generalised Eigenvalue Problems

    E-Print Network [OSTI]

    Wirosoetisno, Djoko

    MA50177: Scientific Computing Case Study Nuclear Reactor Simulation ­ Generalised Eigenvalue of a malfunction or of an accident experimentally, the numerical simulation of nuclear reactors is of utmost balance in a nuclear reactor are the two-group neutron diffusion equations -div (K1 u1) + (a,1 + s) u1 = 1

  9. THE ECONOMICS OF NUCLEAR REACTORS: RENAISSANCE OR RELAPSE?

    E-Print Network [OSTI]

    Laughlin, Robert B.

    THE ECONOMICS OF NUCLEAR REACTORS: RENAISSANCE OR RELAPSE? MARK COOPER SENIOR FELLOW FOR ECONOMIC Findings Approach Hope and Hype vs. Reality in Nuclear Reactor Costs The Economic Cost of Low Carbon. INTRODUCTION 10 A. The Troubling History of Nuclear Reactor Costs B. Purpose and Outline II. THE STRUCTURE

  10. Nuclear reactor alignment plate configuration

    DOE Patents [OSTI]

    Altman, David A; Forsyth, David R; Smith, Richard E; Singleton, Norman R

    2014-01-28T23:59:59.000Z

    An alignment plate that is attached to a core barrel of a pressurized water reactor and fits within slots within a top plate of a lower core shroud and upper core plate to maintain lateral alignment of the reactor internals. The alignment plate is connected to the core barrel through two vertically-spaced dowel pins that extend from the outside surface of the core barrel through a reinforcement pad and into corresponding holes in the alignment plate. Additionally, threaded fasteners are inserted around the perimeter of the reinforcement pad and into the alignment plate to further secure the alignment plate to the core barrel. A fillet weld also is deposited around the perimeter of the reinforcement pad. To accomodate thermal growth between the alignment plate and the core barrel, a gap is left above, below and at both sides of one of the dowel pins in the alignment plate holes through with the dowel pins pass.

  11. Effects of Levels of Automation for Advanced Small Modular Reactors: Impacts on Performance, Workload, and Situation Awareness

    SciTech Connect (OSTI)

    Johanna Oxstrand; Katya Le Blanc

    2014-07-01T23:59:59.000Z

    The Human-Automation Collaboration (HAC) research effort is a part of the Department of Energy (DOE) sponsored Advanced Small Modular Reactor (AdvSMR) program conducted at Idaho National Laboratory (INL). The DOE AdvSMR program focuses on plant design and management, reduction of capital costs as well as plant operations and maintenance costs (O&M), and factory production costs benefits.

  12. CRC handbook of nuclear reactors calculations. Vol. II

    SciTech Connect (OSTI)

    Ronen, Y.

    1986-01-01T23:59:59.000Z

    This handbook breaks down the complex field of nuclear reactor calculations into major steps. Each step presents a detailed analysis of the problems to be solved, the parameters involved, and the elaborate computer programs developed to perform the calculations. This book bridges the gap between nuclear reactor theory and the implementation of that theory, including the problems to be encountered and the level of confidence that should be given to the methods described. Volume II: Monte Carlo Calculations for Nuclear Reactors. In-Core Management of Four Reactor Types. In-Core Management in CANDU-PHW Reactors. Reactor Dynamics. The Theory of Neutron Leakage in Reactor Lattices. Index.

  13. U.S. Department Of Energy Advanced Small Modular Reactor R&D Program: Instrumentation, Controls, and Human-Machine Interface (ICHMI) Pathway

    SciTech Connect (OSTI)

    Holcomb, David Eugene [ORNL; Wood, Richard Thomas [ORNL

    2013-01-01T23:59:59.000Z

    Instrumentation, controls, and human-machine interfaces (ICHMI) are essential enabling technologies that strongly influence nuclear power plant performance and operational costs. The nuclear power industry is currently engaged in a transition from traditional analog-based instrumentation, controls, and human-machine interface systems to implementations employing digital technologies. This transition has primarily occurred in an ad hoc fashion through individual system upgrades at existing plants and has been constrained by licenseability concerns. Although the recent progress in constructing new plants has spurred design of more fully digital plant-wide ICHMI systems, the experience base in the nuclear power application domain is limited. Additionally, development of advanced reactor concepts, such as Generation IV designs and small modular reactors, introduces different plant conditions (e.g., higher temperatures, different coolants, etc.) and unique plant configurations (e.g., multiunit plants with shared systems, balance of plant architectures with reconfigurable co-generation options) that increase the need for enhanced ICHMI capabilities to fully achieve industry goals related to economic competitiveness, safety and reliability, sustainability, and proliferation resistance and physical protection. As a result, significant challenges remain to be addressed to enable the nuclear power industry to complete the transition to safe and comprehensive use of modern ICHMI technology. The U.S. Department of Energy (DOE) has recognized that ICHMI research, development, and demonstration (RD&D) is needed to resolve the technical challenges that may compromise the effective and efficient utilization of modern ICHMI technology and consequently inhibit realization of the benefits offered by expanded utilization of nuclear power. Consequently, several DOE programs have substantial ICHMI RD&D elements within their respective research portfolios. This paper describes current ICHMI research in support of advanced small modular reactors. The objectives that can be achieved through execution of the defined RD&D are to provide optimal technical solutions to critical ICHMI issues, resolve technology gaps arising from the unique measurement and control characteristics of advanced reactor concepts, provide demonstration of needed technologies and methodologies in the nuclear power application domain, mature emerging technologies to facilitate commercialization, and establish necessary technical evidence and application experience to enable timely and predictable licensing. 1 Introduction Instrumentation, controls, and human-machine interfaces are essential enabling technologies that strongly influence nuclear power plant performance and operational costs. The nuclear power industry is currently engaged in a transition from traditional analog-based instrumentation, controls, and human-machine interface (ICHMI) systems to implementations employing digital technologies. This transition has primarily occurred in an ad hoc fashion through individual system upgrades at existing plants and has been constrained by licenseability concerns. Although the recent progress in constructing new plants has spurred design of more fully digital plant-wide ICHMI systems, the experience base in the nuclear power application domain is limited. Additionally, development of advanced reactor concepts, such as Generation IV designs and small modular reactors, introduces different plant conditions (e.g., higher temperatures, different coolants, etc.) and unique plant configurations (e.g., multiunit plants with shared systems, balance of plant architectures with reconfigurable co-generation options) that increase the need for enhanced ICHMI capabilities to fully achieve industry goals related to economic competitiveness, safety and reliability, sustainability, and proliferation resistance and physical protection. As a result, significant challenges remain to be addressed to enable the nuclear power industry to complete the transition to safe and comprehensive use of m

  14. Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Technical Exchange Meeting

    SciTech Connect (OSTI)

    Curtis Smith

    2013-09-01T23:59:59.000Z

    During FY13, the INL developed an advanced SMR PRA framework which has been described in the report Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Detailed Technical Framework Specification, INL/EXT-13-28974 (April 2013). In this framework, the various areas are considered: Probabilistic models to provide information specific to advanced SMRs Representation of specific SMR design issues such as having co-located modules and passive safety features Use of modern open-source and readily available analysis methods Internal and external events resulting in impacts to safety All-hazards considerations Methods to support the identification of design vulnerabilities Mechanistic and probabilistic data needs to support modeling and tools In order to describe this framework more fully and obtain feedback on the proposed approaches, the INL hosted a technical exchange meeting during August 2013. This report describes the outcomes of that meeting.

  15. Nuclear reactor control room construction

    DOE Patents [OSTI]

    Lamuro, R.C.; Orr, R.

    1993-11-16T23:59:59.000Z

    A control room for a nuclear plant is disclosed. In the control room, objects labelled 12, 20, 22, 26, 30 in the drawing are no less than four inches from walls labelled 10.2. A ceiling contains cooling fins that extend downwards toward the floor from metal plates. A concrete slab is poured over the plates. Studs are welded to the plates and are encased in the concrete. 6 figures.

  16. Nuclear reactor control room construction

    DOE Patents [OSTI]

    Lamuro, Robert C. (Pittsburgh, PA); Orr, Richard (Pittsburgh, PA)

    1993-01-01T23:59:59.000Z

    A control room 10 for a nuclear plant is disclosed. In the control room, objects 12, 20, 22, 26, 30 are no less than four inches from walls 10.2. A ceiling 32 contains cooling fins 35 that extend downwards toward the floor from metal plates 34. A concrete slab 33 is poured over the plates. Studs 36 are welded to the plates and are encased in the concrete.

  17. DOE fundamentals handbook: Nuclear physics and reactor theory. Volume 2

    SciTech Connect (OSTI)

    Not Available

    1993-01-01T23:59:59.000Z

    The Nuclear Physics and Reactor Theory Handbook was developed to assist nuclear facility operating contractors in providing operators, maintenance personnel, and the technical staff with the necessary fundamentals training to ensure a basic understanding of nuclear physics and reactor theory. The handbook includes information on atomic and nuclear physics; neutron characteristics; reactor theory and nuclear parameters; and the theory of reactor operation. This information will provide personnel with a foundation for understanding the scientific principles that are associated with various DOE nuclear facility operations and maintenance.

  18. DOE fundamentals handbook: Nuclear physics and reactor theory

    SciTech Connect (OSTI)

    Not Available

    1993-01-01T23:59:59.000Z

    The Nuclear Physics and Reactor Theory Handbook was developed to assist nuclear facility operating contractors in providing operators, maintenance personnel, and the technical staff with the necessary fundamentals training to ensure a basic understanding of nuclear physics and reactor theory. The handbook includes information on atomic and nuclear physics; neutron characteristics; reactor theory and nuclear parameters; and the theory of reactor operation. This information will provide personnel with a foundation for understanding the scientific principles that are associated with various DOE nuclear facility operations and maintenance.

  19. DOE fundamentals handbook: Nuclear physics and reactor theory. Volume 1

    SciTech Connect (OSTI)

    Not Available

    1993-01-01T23:59:59.000Z

    The Nuclear Physics and Reactor Theory Handbook was developed to assist nuclear facility operating contractors in providing operators, maintenance personnel, and the technical staff with the necessary fundamentals training to ensure a basic understanding of nuclear physics and reactor theory. The handbook includes information on atomic and nuclear physics; neutron characteristics; reactor theory and nuclear parameters; and the theory of reactor operation. This information will provide personnel with a foundation for understanding the scientific principles that are associated with various DOE nuclear facility operations and maintenance.

  20. Radioactive target needs for nuclear reactor physics and nuclear astrophysics , G. Barreau1

    E-Print Network [OSTI]

    Boyer, Edmond

    Radioactive target needs for nuclear reactor physics and nuclear astrophysics B.Jurado1* , G sections of short-lived nuclei are key inputs for new generation nuclear reactor simulations and for models and reactor physics where the demand of nuclear data on unstable nuclei is strong, we describe the general

  1. Method for automatically scramming a nuclear reactor

    DOE Patents [OSTI]

    Ougouag, Abderrafi M.; Schultz, Richard R.; Terry, William K.

    2005-12-27T23:59:59.000Z

    An automatically scramming nuclear reactor system. One embodiment comprises a core having a coolant inlet end and a coolant outlet end. A cooling system operatively associated with the core provides coolant to the coolant inlet end and removes heated coolant from the coolant outlet end, thus maintaining a pressure differential therebetween during a normal operating condition of the nuclear reactor system. A guide tube is positioned within the core with a first end of the guide tube in fluid communication with the coolant inlet end of the core, and a second end of the guide tube in fluid communication with the coolant outlet end of the core. A control element is positioned within the guide tube and is movable therein between upper and lower positions, and automatically falls under the action of gravity to the lower position when the pressure differential drops below a safe pressure differential.

  2. CRC handbook of nuclear reactors calculations. Vol. III

    SciTech Connect (OSTI)

    Ronen, Y.

    1986-01-01T23:59:59.000Z

    This handbook breaks down the complex field of nuclear reactor calculations into major steps. Each step presents a detailed analysis of the problems to be solved, the parameters involved, and the elaborate computer programs developed to perform the calculations. This book bridges the gap between nuclear reactor theory and the implementation of that theory, including the problems to be encountered and the level of confidence that should be given to the methods described. Volume III: Control Rods and Burnable Absorber Calculations. Perturbation Theory for Nuclear Reactor Analysis. Thermal Reactors Calculations. Fast Reactor Calculations. Seed-Blanket Reactors. Index.

  3. Oklo reactors and implications for nuclear science

    E-Print Network [OSTI]

    E. D. Davis; C. R. Gould; E. I. Sharapov

    2014-04-19T23:59:59.000Z

    We summarize the nuclear physics interests in the Oklo natural nuclear reactors, focusing particularly on developments over the past two decades. Modeling of the reactors has become increasingly sophisticated, employing Monte Carlo simulations with realistic geometries and materials that can generate both the thermal and epithermal fractions. The water content and the temperatures of the reactors have been uncertain parameters. We discuss recent work pointing to lower temperatures than earlier assumed. Nuclear cross sections are input to all Oklo modeling and we discuss a parameter, the $^{175}$Lu ground state cross section for thermal neutron capture leading to the isomer $^{176\\mathrm{m}}$ Lu, that warrants further investigation. Studies of the time dependence of dimensionless fundamental constants have been a driver for much of the recent work on Oklo. We critically review neutron resonance energy shifts and their dependence on the fine structure constant $\\alpha$ and the ratio $X_q=m_q/\\Lambda$ (where $m_q$ is the average of the $u$ and $d$ current quark masses and $\\Lambda$ is the mass scale of quantum chromodynamics). We suggest a formula for the combined sensitivity to $\\alpha$ and $X_q$ that exhibits the dependence on proton number $Z$ and mass number $A$, potentially allowing quantum electrodynamic and quantum chromodynamic effects to be disentangled if a broader range of isotopic abundance data becomes available.

  4. Optimization of a Small Modular Lead Fast Reactor with Steam Cycle for Remote Siting

    SciTech Connect (OSTI)

    Feldman, Earl E.; Wei, Thomas Y. C.; Sienicki, James J. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, Illinois 60439 (United States)

    2004-07-01T23:59:59.000Z

    Parametric thermal-hydraulic studies needed to develop and optimize the design of a small modular 25 MWt lead-bismuth reactor plant have been performed. The starting point was the design of a liquid metal version of the secure transportable autonomous reactor (STAR-LM) plant of 300 to 400 MWt with a steam power cycle.1 The primary flow is driven entirely by natural convection. The new plant is to be extremely small so that its main components can be transported to the reactor site by truck. The analytical model includes the two major components of the primary loop, the reactor and a once-through steam generator, which is a shell-and-tube heat exchanger with straight vertical tubes. The modeling includes the changes between the beginning and the end of plant life due to the gradual buildup of a layer of magnetite on the surfaces of the fuel pins and on the outer surfaces of the steam generator tubes. Three reactor parametric studies were performed-one for each of three sets of reactor geometric parameters. In each study the pin-bundle pressure drop, the vertical height of the primary loop, the hydraulic diameter of the core, the number of fuel pins, and peak fuel and cladding temperatures were determined for a range of values of fuel pin linear power. Four steam generator parametric studies were performed. The first three have fixed tube inner diameters of 0.5, 1.0, and 1.5 cm, respectively. In the fourth study the tube inner diameter was allowed to vary and the margin to critical heat flux, CHF, was maintained at 20%. In the steam generator studies the independent parameters include tube length and tube-bundle pitch-to-diameter ratio and the dependent variables include steam generator cross-sectional area, the number of tubes, the vertical height of the primary loop, and the steam generator pressure drop. The results show that an acceptable optimum thermal-hydraulic design for a 25 MWt STAR-LM is feasible. (authors)

  5. Thermoacoustic Thermometry for Nuclear Reactor Monitoring

    SciTech Connect (OSTI)

    James A. Smith; Dale K. Kotter; Steven L. Garrett; Randall A. Ali

    2013-06-01T23:59:59.000Z

    On Friday, March 11, 2011, at 2:46pm (Japan Standard Trme), the Tohoku region on the east coast of northern Japan experi­enced what would become known as the largest earthquake in the country's history at magnitude 9.0 on the Richter scale. The Fukushima Daiichi nuclear power plant suffered exten­sive and irreversible damage. Six operating units were at the site, each with a boiling water reactor. When the earthquake struck, three of the six reactors were operating and the others were in a periodic inspection outage phase. In one reactor, all of the fuel had been relocated to a spent fuel pool in the reactor building. The seismic acceleration caused by the earthquake brought the three operating units to an automatic shutdown. Since there was damage to the power transmission lines, the emergency diesel generators (EDG) were automat­ically started to ensure continued cooling of the reactors and spent fuel pools. The situation was under control until the tsunami hit about forty-five minutes later with a maximum wave height of approximately 15 meters, which was three times taller than the sea wall of 5m. The influx of water submerged the EDGs, the electrical switchgear, and dc batteries, resulting in the total loss of power to five of the six reactors. The flooding also resulted in the loss of instrumentation that would have other­ wise been used to monitor and control the emergency. The ugly aftermath included high radiation exposure to operators at the nuclear power plants and early contamina­tion of food supplies and water within several restricted areas in Japan, where high radiation levels have rendered them un­safe for human habitation. While the rest of the story will remain a tragic history, it is this part of the series of unfortunate events that has inspired our research. It has indubitably highlighted the need for a novel sensor and instrumentation system that can withstand similar or worse conditions to avoid future catastrophe and assume damage prevention as quickly as possible. This is the question which we are attempting to answer: Is it possible to implement a self-powered sensor that could transmit data independently of electronic networks while taking advantage of the harsh operating environment of the nuclear reactor?

  6. Uncertainties in the Anti-neutrino Production at Nuclear Reactors

    E-Print Network [OSTI]

    Z. Djurcic; J. A. Detwiler; A. Piepke; V. R. Foster Jr.; L. Miller; G. Gratta

    2008-08-06T23:59:59.000Z

    Anti-neutrino emission rates from nuclear reactors are determined from thermal power measurements and fission rate calculations. The uncertainties in these quantities for commercial power plants and their impact on the calculated interaction rates in electron anti-neutrino detectors is examined. We discuss reactor-to-reactor correlations between the leading uncertainties and their relevance to reactor anti-neutrino experiments.

  7. LIMITED POWER BURSTS IN DISTRIBUTED MODELS OF NUCLEAR REACTORS

    E-Print Network [OSTI]

    Bazhenov, Maxim

    LIMITED POWER BURSTS IN DISTRIBUTED MODELS OF NUCLEAR REACTORS M. V. Bazhenov and E. F. Sabaev UDC employed for analyzing reactor dynamics. Equations of this type are used for analyzing the stability of the reactor power, etc. Among these problems the question of the boundedness of reactor power bursts

  8. INEEL/EXT-01-01623 MODULAR PEBBLE-BED REACTOR PROJECT

    E-Print Network [OSTI]

    for Nuclear Weapons 39 #12;iv 3.1.4.1 Introduction 39 3.1.4.2 Methodology 39 3.1.4.3 Results 40 3 Eigenvalue of HTR Modul 200 and Eskom PBMR 34 3.1.3.2 Support of Planning for Testing of Eskom PBMR Fuel in the Advanced Test Reactor 36 3.1.4 Study of the Potential for PBRs to be Diverted for Production of Material

  9. REPRESENTATIVE SOURCE TERMS AND THE INFLUENCE OF REACTOR ATTRIBUTES ON FUNCTIONAL CONTAINMENT IN MODULAR HIGH-TEMPERATURE GAS-COOLED REACTORS

    SciTech Connect (OSTI)

    Petti, D. A.; Hobbins, R. R.; Lowry, Peter P.; Gougar, Hans

    2013-11-01T23:59:59.000Z

    Modular high-temperature gas-cooled reactors (MHTGRs) offer a high degree of passive safety. The low power density of the reactor and the high heat capacity of the graphite core result in slow transients that do not challenge the integrity of the robust TRISO fuel. Another benefit of this fuel form and the surrounding graphite is their superior ability to retain fission products under all anticipated normal and off-normal conditions, which limits reactor accident source terms to very low values. In this paper, we develop estimates of the source term for a generic MHTGR to illustrate the performance of the radionuclide barriers that comprise the MHTGR functional containment. We also examine the influence of initial fuel quality, fuel performance/failure, reactor outlet temperature, and retention outside of the reactor core on the resultant source term to the environment.

  10. Representative Source Terms and the Influence of Reactor Attributes on Functional Containment in Modular High-Temperature Gas-Cooled Reactors

    SciTech Connect (OSTI)

    D. A. Petti; Hans Gougar; Dick Hobbins; Pete Lowry

    2013-11-01T23:59:59.000Z

    Modular high-temperature gas-cooled reactors (MHTGRs) offer a high degree of passive safety. The low power density of the reactor and the high heat capacity of the graphite core result in slow transients that do not challenge the integrity of the robust TRISO fuel. Another benefit of this fuel form and the surrounding graphite is their superior ability to retain fission products under all anticipated normal and off-normal conditions, which limits reactor accident source terms to very low values. In this paper, we develop estimates of the source term for a generic MHTGR to illustrate the performance of the radionuclide barriers that comprise the MHTGR functional containment. We also examine the influence of initial fuel quality, fuel performance/failure, reactor outlet temperature, and retention outside of the reactor core on the resultant source term to the environment.

  11. High Flux Isotope Reactor named Nuclear Historic Landmark | ornl...

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

    High Flux Isotope Reactor named Nuclear Historic Landmark The High Flux Isotope Reactor vessel at Oak Ridge National Laboratory resides in a pool of water illuminated by the blue...

  12. CRAD, Nuclear Reactor Facility Operations - December 4, 2014...

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

    Reactor Facility Operations - December 4, 2014 (EA CRAD 31-08, Rev. 0) CRAD, Nuclear Reactor Facility Operations - December 4, 2014 (EA CRAD 31-08, Rev. 0) December 4, 2014 CRAD,...

  13. Polynomial regression with derivative information in nuclear reactor uncertainty quantification*

    E-Print Network [OSTI]

    Anitescu, Mihai

    1 Polynomial regression with derivative information in nuclear reactor uncertainty quantification in the outputs. The usual difficulties in modeling the work of the nuclear reactor models include the large size, Argonne National Laboratory, Argonne, IL, USA b Nuclear Engineering Division, Argonne National Laboratory

  14. C Produced by Nuclear Power Reactors Generation and Characterization of

    E-Print Network [OSTI]

    Haviland, David

    14 C Produced by Nuclear Power Reactors ­ Generation and Characterization of Gaseous, Liquid and process water from nuclear reactors ­ A method for quantitative determination of organic and inorganic and Solid Waste �sa Magnusson Division of Nuclear Physics Department of Physics 2007 Akademisk avhandling

  15. Nuclear Thermal Rockets: The Physics of the Fission Reactor

    E-Print Network [OSTI]

    Ross, Shane

    Nuclear Thermal Rockets: The Physics of the Fission Reactor Shane D. Ross Control and Dynamical combustion are those powered by nuclear fission. Comparison of Chemical and Nuclear Rockets. Most existent.g., hydrogen and oxygen). In a nuclear rocket, or more precisely, a nuclear thermal rocket, the propellant

  16. Fuel handling system for a nuclear reactor

    DOE Patents [OSTI]

    Saiveau, James G. (Hickory Hills, IL); Kann, William J. (Park Ridge, IL); Burelbach, James P. (Glen Ellyn, IL)

    1986-01-01T23:59:59.000Z

    A pool type nuclear fission reactor has a core, with a plurality of core elements and a redan which confines coolant as a hot pool at a first end of the core separated from a cold pool at a second end of the core by the redan. A fuel handling system for use with such reactors comprises a core element storage basket located outside of the redan in the cold pool. An access passage is formed in the redan with a gate for opening and closing the passage to maintain the temperature differential between the hot pool and the cold pool. A mechanism is provided for opening and closing the gate. A lifting arm is also provided for manipulating the fuel core elements through the access passage between the storage basket and the core when the redan gate is open.

  17. Closure head for a nuclear reactor

    DOE Patents [OSTI]

    Wade, Elman E. (South Huntingdon, PA)

    1980-01-01T23:59:59.000Z

    A closure head for a nuclear reactor includes a stationary outer ring integral with the reactor vessel with a first rotatable plug disposed within the stationary outer ring and supported from the stationary outer ring by a bearing assembly. A sealing system is associated with the bearing assembly to seal the annulus defined between the first rotatable plug and the stationary outer ring. The sealing system comprises tubular seal elements disposed in the annulus with load springs contacting the tubular seal elements so as to force the tubular seal elements against the annulus in a manner to seal the annulus. The sealing system also comprises a sealing fluid which is pumped through the annulus and over the tubular seal elements causing the load springs to compress thereby reducing the friction between the tubular seal elements and the rotatable components while maintaining a gas-tight seal therebetween.

  18. Nuclear reactor flow control method and apparatus

    DOE Patents [OSTI]

    Church, John P. (1204 Woodbine Rd., Aiken, SC 29803)

    1993-01-01T23:59:59.000Z

    Method and apparatus for improving coolant flow in a nuclear reactor during accident as well as nominal conditions. The reactor has a plurality of fuel elements in sleeves and a plenum above the fuel and through which the sleeves penetrate. Holes are provided in the sleeve so that coolant from the plenum can enter the sleeve and cool the fuel. The number and size of the holes are varied from sleeve to sleeve with the number and size of holes being greater for sleeves toward the center of the core and less for sleeves toward the periphery of the core. Preferably the holes are all the same diameter and arranged in rows and columns, the rows starting from the bottom of every sleeve and fewer rows in peripheral sleeves and more rows in the central sleeves.

  19. Nuclear reactor flow control method and apparatus

    DOE Patents [OSTI]

    Church, J.P.

    1993-03-30T23:59:59.000Z

    Method and apparatus for improving coolant flow in a nuclear reactor during accident as well as nominal conditions. The reactor has a plurality of fuel elements in sleeves and a plenum above the fuel and through which the sleeves penetrate. Holes are provided in the sleeve so that coolant from the plenum can enter the sleeve and cool the fuel. The number and size of the holes are varied from sleeve to sleeve with the number and size of holes being greater for sleeves toward the center of the core and less for sleeves toward the periphery of the core. Preferably the holes are all the same diameter and arranged in rows and columns, the rows starting from the bottom of every sleeve and fewer rows in peripheral sleeves and more rows in the central sleeves.

  20. Neutrino Oscillation Experiments at Nuclear Reactors

    E-Print Network [OSTI]

    Giorgio Gratta

    1999-05-06T23:59:59.000Z

    In this paper I give an overview of the status of neutrino oscillation experiments performed using nuclear reactors as sources of neutrinos. I review the present generation of experiments (Chooz and Palo Verde) with baselines of about 1 km as well as the next generation that will search for oscillations with a baseline of about 100 km. While the present detectors provide essential input towards the understanding of the atmospheric neutrino anomaly, in the future, the KamLAND reactor experiment represents our best opportunity to study very small mass neutrino mixing in laboratory conditions. In addition KamLAND with its very large fiducial mass and low energy threshold, will also be sensitive to a broad range of different physics.

  1. Monthly Nuclear Utility Generation by State and Reactor, 2007

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

    applicationvnd.ms-excel X-Translator-Status: translating " Worksheet" "Monthly Nuclear Utility Generation by State and Reactor, 2007" "January through December 2007"...

  2. Monthly Nuclear Utility Generation by State and Reactor, 2004

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

    applicationvnd.ms-excel X-Translator-Status: translating " Worksheet" "Monthly Nuclear Utility Generation by State and Reactor, 2004" "January through December 2004"...

  3. Monthly Nuclear Utility Generation by State and Reactor, 2005

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

    applicationvnd.ms-excel X-Translator-Status: translating " Worksheet" "Monthly Nuclear Utility Generation by State and Reactor, 2005" "January through December 2005"...

  4. Monthly Nuclear Utility Generation by State and Reactor, 2003

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

    applicationvnd.ms-excel X-Translator-Status: translating " Worksheet" "Monthly Nuclear Utility Generation by State and Reactor, 2003" "January through December 2003"...

  5. Monthly Nuclear Utility Generation by State and Reactor, 2008

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

    applicationvnd.ms-excel X-Translator-Status: translating " Worksheet" "Monthly Nuclear Utility Generation by State and Reactor, 2008" "January through December 2008"...

  6. Discovery sheds light on nuclear reactor fuel behavior during...

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

    and Reports Summer Science Writing Internship Discovery sheds light on nuclear reactor fuel behavior during a severe event By Angela Hardin * November 20, 2014 Tweet...

  7. Report of the Nuclear Reactor Technology Subcommittee

    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: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn April 23, 2014, an OHASeptemberAssessments | Department ofSouthernof the Nuclear Reactor

  8. Identification of Selected Areas to Support Federal Clean Energy Goals Using Small Modular Reactors

    SciTech Connect (OSTI)

    Belles, R. J. [ORNL; Mays, G. T. [ORNL; Omitaomu, O. A. [ORNL; Poore, W. P. [ORNL

    2013-12-30T23:59:59.000Z

    Beginning in late 2008, Oak Ridge National Laboratory (ORNL) responded to ongoing internal and external studies addressing key questions related to our national electrical energy supply. This effort has led to the development and refinement of Oak Ridge Siting Analysis for power Generation Expansion (OR-SAGE), a tool to support power plant siting evaluations. The objective in developing OR-SAGE was to use industry-accepted approaches and/or develop appropriate criteria for screening sites and employ an array of geographic information systems (GIS) data sources at ORNL to identify candidate areas for a power generation technology application. The basic premise requires the development of exclusionary, avoidance, and suitability criteria for evaluating sites for a given siting application, such as siting small modular reactors (SMRs). For specific applications of the tool, it is necessary to develop site selection and evaluation criteria (SSEC) that encompass a number of key benchmarks that essentially form the site environmental characterization for that application. These SSEC might include population density, seismic activity, proximity to water sources, proximity to hazardous facilities, avoidance of protected lands and floodplains, susceptibility to landslide hazards, and others.

  9. Fluid sampling system for a nuclear reactor

    DOE Patents [OSTI]

    Lau, Louis K. (Monroeville, PA); Alper, Naum I. (Monroeville, PA)

    1994-01-01T23:59:59.000Z

    A system of extracting fluid samples, either liquid or gas, from the interior of a nuclear reactor containment utilizes a jet pump. To extract the sample fluid, a nonradioactive motive fluid is forced through the inlet and discharge ports of a jet pump located outside the containment, creating a suction that draws the sample fluid from the containment through a sample conduit connected to the pump suction port. The mixture of motive fluid and sample fluid is discharged through a return conduit to the interior of the containment. The jet pump and means for removing a portion of the sample fluid from the sample conduit can be located in a shielded sample grab station located next to the containment. A non-nuclear grade active pump can be located outside the grab sampling station and the containment to pump the nonradioactive motive fluid through the jet pump.

  10. Fluid sampling system for a nuclear reactor

    DOE Patents [OSTI]

    Lau, L.K.; Alper, N.I.

    1994-11-22T23:59:59.000Z

    A system of extracting fluid samples, either liquid or gas, from the interior of a nuclear reactor containment utilizes a jet pump. To extract the sample fluid, a nonradioactive motive fluid is forced through the inlet and discharge ports of a jet pump located outside the containment, creating a suction that draws the sample fluid from the containment through a sample conduit connected to the pump suction port. The mixture of motive fluid and sample fluid is discharged through a return conduit to the interior of the containment. The jet pump and means for removing a portion of the sample fluid from the sample conduit can be located in a shielded sample grab station located next to the containment. A non-nuclear grade active pump can be located outside the grab sampling station and the containment to pump the nonradioactive motive fluid through the jet pump. 1 fig.

  11. Direct conversion nuclear reactor space power systems

    SciTech Connect (OSTI)

    Britt, E.J.; Fitzpatrick, G.O.

    1982-08-01T23:59:59.000Z

    This paper presents the results of a study of space nuclear reactor power systems using either thermoelectric or thermionic energy converters. An in-core reactor design and two heat pipe cooled out-of-core reactor designs were considered. One of the out-of-core cases utilized, long heat pipes (LHP) directly coupled to the energy converter. The second utilized a larger number of smaller heat pipes (mini-pipe) radiatively coupled to the energy converter. In all cases the entire system, including power conditioning, was constrained to be launched in a single shuttle flight. Assuming presently available performance, both the LHP thermoelectric system and minipipe thermionic system, designed to produce 100 kWe for seven years, would have a specific mass near 22kg/kWe. The specific mass of the thermionic minipipe system designed for a one year mission is 165 kg/kWe due to less fuel swelling. Shuttle imposed growth limits are near 300 kWe and 1.2 MWe for the thermoelectric and thermionic systems, respectively. Converter performance improvements could double this potential, and over 10 MWe may be possible for very short missions.

  12. Electrochemistry of Water-Cooled Nuclear Reactors

    SciTech Connect (OSTI)

    Macdonald, Dgiby; Urquidi-Macdonald, Mirna; Pitt, Jonathan

    2006-08-08T23:59:59.000Z

    This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or "radiation fields" around the primary loop and the vessel, as a function of the operating parameters and the water chemistry.

  13. Neutron transport analysis for nuclear reactor design

    DOE Patents [OSTI]

    Vujic, Jasmina L. (Lisle, IL)

    1993-01-01T23:59:59.000Z

    Replacing regular mesh-dependent ray tracing modules in a collision/transfer probability (CTP) code with a ray tracing module based upon combinatorial geometry of a modified geometrical module (GMC) provides a general geometry transfer theory code in two dimensions (2D) for analyzing nuclear reactor design and control. The primary modification of the GMC module involves generation of a fixed inner frame and a rotating outer frame, where the inner frame contains all reactor regions of interest, e.g., part of a reactor assembly, an assembly, or several assemblies, and the outer frame, with a set of parallel equidistant rays (lines) attached to it, rotates around the inner frame. The modified GMC module allows for determining for each parallel ray (line), the intersections with zone boundaries, the path length between the intersections, the total number of zones on a track, the zone and medium numbers, and the intersections with the outer surface, which parameters may be used in the CTP code to calculate collision/transfer probability and cross-section values.

  14. Neutron transport analysis for nuclear reactor design

    DOE Patents [OSTI]

    Vujic, J.L.

    1993-11-30T23:59:59.000Z

    Replacing regular mesh-dependent ray tracing modules in a collision/transfer probability (CTP) code with a ray tracing module based upon combinatorial geometry of a modified geometrical module (GMC) provides a general geometry transfer theory code in two dimensions (2D) for analyzing nuclear reactor design and control. The primary modification of the GMC module involves generation of a fixed inner frame and a rotating outer frame, where the inner frame contains all reactor regions of interest, e.g., part of a reactor assembly, an assembly, or several assemblies, and the outer frame, with a set of parallel equidistant rays (lines) attached to it, rotates around the inner frame. The modified GMC module allows for determining for each parallel ray (line), the intersections with zone boundaries, the path length between the intersections, the total number of zones on a track, the zone and medium numbers, and the intersections with the outer surface, which parameters may be used in the CTP code to calculate collision/transfer probability and cross-section values. 28 figures.

  15. Minimizing or eliminating refueling of nuclear reactor

    DOE Patents [OSTI]

    Doncals, Richard A. (Washington, PA); Paik, Nam-Chin (Pittsburgh, PA); Andre, Sandra V. (Hempfield Township, Westmoreland County, PA); Porter, Charles A. (Rostraver Township, Westmoreland County, PA); Rathbun, Roy W. (Greensburg, PA); Schwallie, Ambrose L. (Greensburg, PA); Petras, Diane S. (Penn Township, Westmoreland County, PA)

    1989-01-01T23:59:59.000Z

    Demand for refueling of a liquid metal fast nuclear reactor having a life of 30 years is eliminated or reduced to intervals of at least 10 years by operating the reactor at a low linear-power density, typically 2.5 kw/ft of fuel rod, rather than 7.5 or 15 kw/ft, which is the prior art practice. So that power of the same magnitude as for prior art reactors is produced, the volume of the core is increased. In addition, the height of the core and it diameter are dimensioned so that the ratio of the height to the diameter approximates 1 to the extent practicable considering the requirement of control and that the pressure drop in the coolant shall not be excessive. The surface area of a cylinder of given volume is a minimum if the ratio of the height to the diameter is 1. By minimizing the surface area, the leakage of neutrons is reduced. By reducing the linear-power density, increasing core volume, reducing fissile enrichment and optimizing core geometry, internal-core breeding of fissionable fuel is substantially enhanced. As a result, core operational life, limited by control worth requirements and fuel burnup capability, is extended up to 30 years of continuous power operation.

  16. Nuclear reactors built, being built, or planned 1993

    SciTech Connect (OSTI)

    Not Available

    1993-08-01T23:59:59.000Z

    Nuclear Reactors Built, Being Built, or Planned contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1993. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: civilian, production, military, export, and critical assembly.

  17. Spent nuclear fuel discharges from U.S. reactors 1994

    SciTech Connect (OSTI)

    NONE

    1996-02-01T23:59:59.000Z

    Spent Nuclear Fuel Discharges from US Reactors 1994 provides current statistical data on fuel assemblies irradiated at commercial nuclear reactors operating in the US. This year`s report provides data on the current inventories and storage capacities at these reactors. Detailed statistics on the data are presented in four chapters that highlight 1994 spent fuel discharges, storage capacities and inventories, canister and nonfuel component data, and assembly characteristics. Five appendices, a glossary, and bibliography are also included. 10 figs., 34 tabs.

  18. Uncertainties in the Anti-neutrino Production at Nuclear Reactors

    SciTech Connect (OSTI)

    Djurcic, Zelimir; Detwiler, Jason A.; Piepke, Andreas; Foster Jr., Vince R.; Miller, Lester; Gratta, Giorgio

    2008-08-06T23:59:59.000Z

    Anti-neutrino emission rates from nuclear reactors are determined from thermal power measurements and fission rate calculations. The uncertainties in these quantities for commercial power plants and their impact on the calculated interaction rates in {bar {nu}}{sub e} detectors is examined. We discuss reactor-to-reactor correlations between the leading uncertainties, and their relevance to reactor {bar {nu}}{sub e} experiments.

  19. Weld monitor and failure detector for nuclear reactor system

    DOE Patents [OSTI]

    Sutton, Jr., Harry G. (Mt. Lebanon, PA)

    1987-01-01T23:59:59.000Z

    Critical but inaccessible welds in a nuclear reactor system are monitored throughout the life of the reactor by providing small aperture means projecting completely through the reactor vessel wall and also through the weld or welds to be monitored. The aperture means is normally sealed from the atmosphere within the reactor. Any incipient failure or cracking of the weld will cause the environment contained within the reactor to pass into the aperture means and thence to the outer surface of the reactor vessel where its presence is readily detected.

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

    SciTech Connect (OSTI)

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

    2007-03-21T23:59:59.000Z

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

  1. Nuclear reactor cooling system decontamination reagent regeneration

    DOE Patents [OSTI]

    Anstine, Larry D. (San Jose, CA); James, Dean B. (Saratoga, CA); Melaika, Edward A. (Berkeley, CA); Peterson, Jr., John P. (Livermore, CA)

    1985-01-01T23:59:59.000Z

    An improved method for decontaminating the coolant system of water-cooled nuclear power reactors and for regenerating the decontamination solution. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution.

  2. Design of a nuclear reactor system for lunar base applications

    E-Print Network [OSTI]

    Griffith, Richard Odell

    1986-01-01T23:59:59.000Z

    DESIGN OF A NUCLEAR REACTOR SYSTEM FOR LUNAR BASE APPLICATIONS A Thesis by RICMARD ODELL GRIFFITH Submitted to the Graduate College of Texas A&M University in partial fulfillment of the requirements for the degree of MASTER OF SCIENCE... August 1986 Major Subject: Nuclear Engineer ing DESIGN OF A NUCLEAR REACTOR SYSTEM FOR LUNAR BASE APPLICATIONS A Thesis by RICHARD ODELL GRIFFITH Appr oved as to style and content by: Carl A. Endman (Chain of Committee) Cer aid A. Schlapper...

  3. MOOSE simulating nuclear reactor CRUD buildup

    ScienceCinema (OSTI)

    None

    2014-07-21T23:59:59.000Z

    This simulation uses multiple physical models to show how the buildup of boron deposits on reactor fuel can affect performance and the reactor's power profile.

  4. MOOSE simulating nuclear reactor CRUD buildup

    SciTech Connect (OSTI)

    None

    2014-02-06T23:59:59.000Z

    This simulation uses multiple physical models to show how the buildup of boron deposits on reactor fuel can affect performance and the reactor's power profile.

  5. A brief history of design studies on innovative nuclear reactors

    SciTech Connect (OSTI)

    Sekimoto, Hiroshi, E-mail: hsekimot@gmail.com [Emeritus Professor, Tokyo Institute of Technology (Japan)

    2014-09-30T23:59:59.000Z

    In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970’s the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980’s the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

  6. PHYSICS OF NUCLEAR REACTORS Nuclear reactions and cross sections 1-10

    E-Print Network [OSTI]

    Danon, Yaron

    PHYSICS OF NUCLEAR REACTORS Nuclear reactions and cross sections 1-10 10 11 12 13 14 15 16 17 18 19 Figure 5 ­ neutron capture, fission and elastic scattering in 235 U. For spaced resonances the center neutron wavelength, D is given by: cE mM Mm 2 + = h D , (1.22) 1 Bell and Glasstone, Nuclear Reactor

  7. Nuclear reactors built, being built, or planned, 1991

    SciTech Connect (OSTI)

    Simpson, B.

    1992-07-01T23:59:59.000Z

    This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1991. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is an American company -- working either independently or in cooperation with a foreign company (Part 4, in each section). Critical assembly refers to an assembly of fuel and assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  8. Plutonium Discharge Rates and Spent Nuclear Fuel Inventory Estimates for Nuclear Reactors Worldwide

    SciTech Connect (OSTI)

    Brian K. Castle; Shauna A. Hoiland; Richard A. Rankin; James W. Sterbentz

    2012-09-01T23:59:59.000Z

    This report presents a preliminary survey and analysis of the five primary types of commercial nuclear power reactors currently in use around the world. Plutonium mass discharge rates from the reactors’ spent fuel at reload are estimated based on a simple methodology that is able to use limited reactor burnup and operational characteristics collected from a variety of public domain sources. Selected commercial reactor operating and nuclear core characteristics are also given for each reactor type. In addition to the worldwide commercial reactors survey, a materials test reactor survey was conducted to identify reactors of this type with a significant core power rating. Over 100 material or research reactors with a core power rating >1 MW fall into this category. Fuel characteristics and spent fuel inventories for these material test reactors are also provided herein.

  9. Role of nuclear reactors in future military satellites

    SciTech Connect (OSTI)

    Buden, D.; Angelo, J.A. Jr.

    1982-01-01T23:59:59.000Z

    Future military capabilities will be profoundly influenced by emerging Shuttle Era space technology. Regardless of the specific direction or content of tomorrow's military space program, it is clear that advanced space transportation systems, orbital support facilities, and large-capacity power subsystems will be needed to create the generally larger, more sophisticated military space systems of the future. This paper explores the critical role that space nuclear reactors should play in America's future space program and reviews the current state of nuclear reactor power plant technology. Space nuclear reactor technologies have the potential of satisfying power requirements ranging from 10 kW/sub (e)/ to 100 MW/sub (e)/.

  10. The behavior of fission products during nuclear rocket reactor tests

    SciTech Connect (OSTI)

    Bokor, P.C.; Kirk, W.L.; Bohl, R.J.

    1991-01-01T23:59:59.000Z

    The experience base regarding fission product behavior developed during the Rover program, the nuclear rocket development program of 1955--1972, will be useful in planning a renewed nuclear rocket program. During the Rover program, 20 reactors were tested at the Nuclear Rocket Development Station in Nevada. Nineteen of these discharged effluent directly into the atmosphere; the last reactor tested, a non-flight-prototypic, fuel-element-testing reactor called the Nuclear Furnace (NF-1) was connected to an effluent cleanup system that removed fission products before the hydrogen coolant (propellant) was discharged to the atmosphere. In general, we are able to increase both test duration and fuel temperature during the test series. Therefore fission product data from the later part of the program are more interesting and more applicable to future reactors. We have collected fission product retention (and release) data reported in both formal and informal publications for six of the later reactor tests; five of these were Los Alamos reactors that were firsts of a kind in configuration or operating conditions. We have also, with the cooperation of Westinghouse, included fission product data from the NRX-A6 reactor, the final member of series of developmental reactors with the same basic geometry, but with significant design and fabrication improvements as the series continued. Table 1 lists the six selected reactors and the test parameters for each.

  11. Indirect passive cooling system for liquid metal cooled nuclear reactors

    DOE Patents [OSTI]

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

    1990-01-01T23:59:59.000Z

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

  12. Passive cooling safety system for liquid metal cooled nuclear reactors

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA); Hui, Marvin M. (Sunnyvale, CA); Berglund, Robert C. (Saratoga, CA)

    1991-01-01T23:59:59.000Z

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

  13. INITIATORS AND TRIGGERING CONDITIONS FOR ADAPTIVE AUTOMATION IN ADVANCED SMALL MODULAR REACTORS

    SciTech Connect (OSTI)

    Katya L Le Blanc; Johanna h Oxstrand

    2014-04-01T23:59:59.000Z

    It is anticipated that Advanced Small Modular Reactors (AdvSMRs) will employ high degrees of automation. High levels of automation can enhance system performance, but often at the cost of reduced human performance. Automation can lead to human out-of the loop issues, unbalanced workload, complacency, and other problems if it is not designed properly. Researchers have proposed adaptive automation (defined as dynamic or flexible allocation of functions) as a way to get the benefits of higher levels of automation without the human performance costs. Adaptive automation has the potential to balance operator workload and enhance operator situation awareness by allocating functions to the operators in a way that is sensitive to overall workload and capabilities at the time of operation. However, there still a number of questions regarding how to effectively design adaptive automation to achieve that potential. One of those questions is related to how to initiate (or trigger) a shift in automation in order to provide maximal sensitivity to operator needs without introducing undesirable consequences (such as unpredictable mode changes). Several triggering mechanisms for shifts in adaptive automation have been proposed including: operator initiated, critical events, performance-based, physiological measurement, model-based, and hybrid methods. As part of a larger project to develop design guidance for human-automation collaboration in AdvSMRs, researchers at Idaho National Laboratory have investigated the effectiveness and applicability of each of these triggering mechanisms in the context of AdvSMR. Researchers reviewed the empirical literature on adaptive automation and assessed each triggering mechanism based on the human-system performance consequences of employing that mechanism. Researchers also assessed the practicality and feasibility of using the mechanism in the context of an AdvSMR control room. Results indicate that there are tradeoffs associated with each mechanism, but that some are more applicable to the AdvSMR domain. The two mechanisms that consistently improve performance in laboratory studies are operator initiated adaptive automation based on hierarchical task delegation and the Electroencephalogram(EEG) –based measure of engagement. Current EEG methods are intrusive and require intensive analysis; therefore it is not recommended for an AdvSMR control rooms at this time. Researchers also discuss limitations in the existing empirical literature and make recommendations for further research.

  14. Scaling Studies for High Temperature Test Facility and Modular High Temperature Gas-Cooled Reactor

    SciTech Connect (OSTI)

    Richard R. Schult; Paul D. Bayless; Richard W. Johnson; James R. Wolf; Brian Woods

    2012-02-01T23:59:59.000Z

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5-year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant (NGNP) project. Because the NRC's interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC).

  15. Nuclear reactors built, being built, or planned 1996

    SciTech Connect (OSTI)

    NONE

    1997-08-01T23:59:59.000Z

    This publication contains unclassified information about facilities, built, being built, or planned in the United States for domestic use or export as of December 31, 1996. The Office of Scientific and Technical Information, U.S. Department of Energy, gathers this information annually from Washington headquarters, and field offices of DOE; from the U.S. Nuclear Regulatory Commission (NRC); from the U. S. reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from U.S. and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled.

  16. 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

    2010-01-01T23:59:59.000Z

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

  17. Valuing modular nuclear power plants in finite time decision horizon Shashi Jain a,b,

    E-Print Network [OSTI]

    Oosterlee, Cornelis W. "Kees"

    into account the value of flexibility arising due to modular construction, which traditional valuation methods

  18. Nuclear Graphite -Fission Reactor Brief Outline of Experience and

    E-Print Network [OSTI]

    McDonald, Kirk

    Nuclear Graphite - Fission Reactor Brief Outline of Experience and Understanding Professor Barry J · Physical Changes ­ to Polycrystalline Graphite due to Fast Neutron Damage and Radiolytic Oxidation ­ Provided channels for control rods and coolant gas · Neutron Shield ­ Boronated graphite · Thermal columns

  19. Observer-based fault detection for nuclear reactors

    E-Print Network [OSTI]

    Li, Qing, 1972-

    2001-01-01T23:59:59.000Z

    This is a study of fault detection for nuclear reactor systems. Basic concepts are derived from fundamental theories on system observers. Different types of fault- actuator fault, sensor fault, and system dynamics fault ...

  20. Fuel assembly transfer basket for pool type nuclear reactor vessels

    DOE Patents [OSTI]

    Fanning, Alan W. (San Jose, CA); Ramsour, Nicholas L. (San Jose, CA)

    1991-01-01T23:59:59.000Z

    A fuel assembly transfer basket for a pool type, liquid metal cooled nuclear reactor having a side access loading and unloading port for receiving and relinquishing fuel assemblies during transfer.

  1. Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

    SciTech Connect (OSTI)

    Ehud Greenspan

    2008-09-30T23:59:59.000Z

    This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

  2. Liquid metal cooled nuclear reactors with passive cooling system

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Fanning, Alan W. (San Jose, CA)

    1991-01-01T23:59:59.000Z

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of cooling medium flow circuits which cooperate to remove and carry heat away from the fuel core upon loss of the normal cooling flow circuit to areas external thereto.

  3. Nuclear reactors built, being built, or planned, 1994

    SciTech Connect (OSTI)

    NONE

    1995-07-01T23:59:59.000Z

    This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1994. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; tables of data for reactors operating, being built, or planned; and tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company -- working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  4. Nuclear reactors built, being built, or planned: 1995

    SciTech Connect (OSTI)

    NONE

    1996-08-01T23:59:59.000Z

    This report contains unclassified information about facilities built, being built, or planned in the US for domestic use or export as of December 31, 1995. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company--working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  5. CRC handbook of nuclear reactors calculations. Vol. I

    SciTech Connect (OSTI)

    Ronen, Y.

    1986-01-01T23:59:59.000Z

    This handbook breaks down the complex field of nuclear reactor calculations into major steps. Each step presents a detailed analysis of the problems to be solved, the parameters involved, and the elaborate computer programs developed to perform the calculations. This book bridges the gap between nuclear reactor theory and the implementation of that theory, including the problems to be encountered and the level of confidence that should be given to the methods described.

  6. Neutron spectrometer for fast nuclear reactors

    E-Print Network [OSTI]

    Osipenko, M; Ricco, G; Caiffi, B; Pompili, F; Pillon, M; Angelone, M; Verona-Rinati, G; Cardarelli, R; Mila, G; Argiro, S

    2015-01-01T23:59:59.000Z

    In this paper we describe the development and first tests of a neutron spectrometer designed for high flux environments, such as the ones found in fast nuclear reactors. The spectrometer is based on the conversion of neutrons impinging on $^6$Li into $\\alpha$ and $t$ whose total energy comprises the initial neutron energy and the reaction $Q$-value. The $^6$LiF layer is sandwiched between two CVD diamond detectors, which measure the two reaction products in coincidence. The spectrometer was calibrated at two neutron energies in well known thermal and 3 MeV neutron fluxes. The measured neutron detection efficiency varies from 4.2$\\times 10^{-4}$ to 3.5$\\times 10^{-8}$ for thermal and 3 MeV neutrons, respectively. These values are in agreement with Geant4 simulations and close to simple estimates based on the knowledge of the $^6$Li(n,$\\alpha$)$t$ cross section. The energy resolution of the spectrometer was found to be better than 100 keV when using 5 m cables between the detector and the preamplifiers.

  7. Neutron spectrometer for fast nuclear reactors

    E-Print Network [OSTI]

    M. Osipenko; M. Ripani; G. Ricco; B. Caiffi; F. Pompili; M. Pillon; M. Angelone; G. Verona-Rinati; R. Cardarelli; G. Mila; S. Argiro

    2015-05-25T23:59:59.000Z

    In this paper we describe the development and first tests of a neutron spectrometer designed for high flux environments, such as the ones found in fast nuclear reactors. The spectrometer is based on the conversion of neutrons impinging on $^6$Li into $\\alpha$ and $t$ whose total energy comprises the initial neutron energy and the reaction $Q$-value. The $^6$LiF layer is sandwiched between two CVD diamond detectors, which measure the two reaction products in coincidence. The spectrometer was calibrated at two neutron energies in well known thermal and 3 MeV neutron fluxes. The measured neutron detection efficiency varies from 4.2$\\times 10^{-4}$ to 3.5$\\times 10^{-8}$ for thermal and 3 MeV neutrons, respectively. These values are in agreement with Geant4 simulations and close to simple estimates based on the knowledge of the $^6$Li(n,$\\alpha$)$t$ cross section. The energy resolution of the spectrometer was found to be better than 100 keV when using 5 m cables between the detector and the preamplifiers.

  8. Improved Design of Nuclear Reactor Control System | U.S. DOE...

    Office of Science (SC) Website

    Improved Design of Nuclear Reactor Control System Nuclear Physics (NP) NP Home About Research Facilities Science Highlights Benefits of NP Applications of Nuclear Science...

  9. Fukushima Nuclear Crisis Recovery: A Modular Water Treatment System Deployed in Seven Weeks - 12489

    SciTech Connect (OSTI)

    Denton, Mark S.; Mertz, Joshua L. [Kurion, Inc., P.O. Box 5901, Oak Ridge, Tennessee 37831 (United States); Bostick, William D. [Materials and Chemistry Laboratory, Inc. (MCL) ETTP, Building K-1006, 2010 Highway 58, Suite 1000, Oak Ridge, Tennessee 37830 (United States)

    2012-07-01T23:59:59.000Z

    On March 11, 2011, the magnitude 9.0 Great East Japan earthquake, Tohoku, hit off the Fukushima coast of Japan. This was one of the most powerful earthquakes in recorded history and the most powerful one known to have hit Japan. The ensuing tsunami devastated a huge area resulting in some 25,000 persons confirmed dead or missing. The perfect storm was complete when the tsunami then found the four reactor, Fukushima-Daiichi Nuclear Station directly in its destructive path. While recovery systems admirably survived the powerful earthquake, the seawater from the tsunami knocked the emergency cooling systems out and did extensive damage to the plant and site. Subsequent hydrogen generation caused explosions which extended this damage to a new level and further flooded the buildings with highly contaminated water. Some 2 million people were evacuated from a fifty mile radius of the area and evaluation and cleanup began. Teams were assembled in Tokyo the first week of April to lay out potential plans for the immediate treatment of some 63 million gallons (a number which later exceeded 110 million gallons) of highly contaminated water to avoid overflow from the buildings as well as supply the desperately needed clean cooling water for the reactors. A system had to be deployed with a very brief cold shake down and hot startup before the rainy season started in early June. Joined by team members Toshiba (oil removal system), AREVA (chemical precipitation system) and Hitachi-GE (RO system), Kurion (cesium removal system following the oil separator) proposed, designed, fabricated, delivered and started up a one of a kind treatment skid and over 100 metric tons of specially engineered and modified Ion Specific Media (ISM) customized for this very challenging seawater/oil application, all in seven weeks. After a very short cold shake down, the system went into operation on June 17, 2011 on actual waste waters far exceeding 1 million Bq/mL in cesium and many other isotopes. One must remember that, in addition to attempting to do isotope removal in the competition of seawater (as high as 18,000 ppm sodium due to concentration), some 350,000 gallons of turbine oil was dispersed into the flooded buildings as well. The proposed system consisted of a 4 guard vessel skid for the oil and debris, 4 skids containing 16 cesium towers in a lead-lag layout with removable vessels (sent to an interim storage facility), and a 4 polishing vessel skid for iodine removal and trace cesium levels. At a flow rate of at least 220 gallons per minute, the system has routinely removed over 99% of the cesium, the main component of the activity, since going on line. To date, some 50% of the original activity has been removed and stabilized and cold shutdown of the plant was announced on December 10, 2011. In March and April alone, 10 cubic feet of Engineered Herschelite was shipped to Seabrook Nuclear Power Plant, NPP, to support the April 1, 2011 outage cleanup; 400 cubic feet was shipped to Oak Ridge National Laboratory (ORNL) for strontium (Sr-90) ground water remediation; and 6000 cubic feet (100 metric tons, MT, or 220,400 pounds) was readied for the Fukushima Nuclear Power Station with an additional 100 MT on standby for replacement vessels. This experience and accelerated media production in the U.S. bore direct application to what was to soon be used in Fukushima. How such a sophisticated and totally unique system and huge amount of media could be deployable in such a challenging and changing matrix, and in only seven weeks, is outlined in this paper as well as the system and operation itself. As demonstrated herein, all ten major steps leading up to the readiness and acceptance of a modular emergency technology recovery system were met and in a very short period of time, thus utilizing three decades of experience to produce and deliver such a system literally in seven weeks: - EPRI - U.S. Testing and Experience Leading to Introduction to EPRI - Japan and Subsequently TEPCO Emergency Meetings - Three Mile Island (TMI) Media and Vitrification Experience

  10. Temperature dependent scattering cross section effects on nuclear reactor control

    E-Print Network [OSTI]

    Biggs, Charles Leon

    1968-01-01T23:59:59.000Z

    Reactor e e o e o a a e e ~ a o e ~ a e o o 43 Ato. . . Dsnsitiec of 'Materials in the Conceived Fast Nuclear Reactor ~ ~ o e o e e a o a o e e o ~ ~ ~ ~ 6, IDPut SPecifications oi' the AIYi-6 Criticality arch e o o e a a ~ a e e o ~ ~ ~ e e e ~ o e... both reactors depended upon axially expanding fuel elements for inherent control, other methods should be considered, Due to ths magnitude and sign of the reactivity cosfi'ioients, inherent control is especially of. interest in large fast nuclear...

  11. Evaluating Russian space nuclear reactor technology for United States applications

    SciTech Connect (OSTI)

    Polansky, G.F. [Phillips Lab., Albuquerque, NM (United States); Schmidt, G.L. [New Mexico Engineering Research Institute, Albuquerque, NM (United States); Voss, S.S. [Los Alamos National Lab., NM (United States); Reynolds, E.L. [Applied Physics Lab., Laurel, MD (United States)

    1994-08-01T23:59:59.000Z

    Space nuclear power and nuclear electric propulsion are considered important technologies for planetary exploration, as well as selected earth orbit applications. The Nuclear Electric Propulsion Space Test Program (NEPSTP) was intended to provide an early flight demonstration of these technologies at relatively low cost through extensive use of existing Russian technology. The key element of Russian technology employed in the program was the Topaz II reactor. Refocusing of the activities of the Ballistic Missile Defense Organization (BMDO), combined with budgetary pressures, forced the cancellation of the NEPSTP at the end of the 1993 fiscal year. The NEPSTP was faced with many unique flight qualification issues. In general, the launch of a spacecraft employing a nuclear reactor power system complicates many spacecraft qualification activities. However, the NEPSTP activities were further complicated because the reactor power system was a Russian design. Therefore, this program considered not only the unique flight qualification issues associated with space nuclear power, but also with differences between Russian and United States flight qualification procedures. This paper presents an overview of the NEPSTP. The program goals, the proposed mission, the spacecraft, and the Topaz II space nuclear power system are described. The subject of flight qualification is examined and the inherent difficulties of qualifying a space reactor are described. The differences between United States and Russian flight qualification procedures are explored. A plan is then described that was developed to determine an appropriate flight qualification program for the Topaz II reactor to support a possible NEPSTP launch.

  12. Spent nuclear fuel discharges from US reactors 1992

    SciTech Connect (OSTI)

    Not Available

    1994-05-05T23:59:59.000Z

    This report provides current statistical data on every fuel assembly irradiated in commercial nuclear reactors operating in the United States. It also provides data on the current inventories and storage capacities of those reactors to a wide audience, including Congress, Federal and State agencies, the nuclear and electric industries and the general public. It uses data from the mandatory, ``Nuclear Fuel Data`` survey, Form RW-859 for 1992 and historical data collected by the Energy Information Administration (EIA) on previous Form RW-859 surveys. The report was prepared by the EIA under a Memorandum of Understanding with the Office of Civilian Radioactive Waste Management.

  13. Technology, Safety and Costs of Decommissioning Nuclear Reactors At Multiple-Reactor Stations

    SciTech Connect (OSTI)

    Wittenbrock, N. G.

    1982-01-01T23:59:59.000Z

    Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWRs) and large (1155-MWe) boiling water reactors {BWRs) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite nuclear waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services. Five scenarios for decommissioning reactors at a multiple-reactor station are investigated. The number of reactors on a site is assumed to be either four or ten; nuclear waste disposal is varied between immediate offsite disposal, interim onsite storage, and immediate onsite disposal. It is assumed that the decommissioned reactors are not replaced in one scenario but are replaced in the other scenarios. Centralized service facilities are provided in two scenarios but are not provided in the other three. Decommissioning of a PWR or a BWR at a multiple-reactor station probably will be less costly and result in lower radiation doses than decommissioning an identical reactor at a single-reactor station. Regardless of whether the light water reactor being decommissioned is at a single- or multiple-reactor station: • the estimated occupational radiation dose for decommissioning an LWR is lowest for SAFSTOR and highest for DECON • the estimated cost of decommissioning a PWR is lowest for ENTOMB and highest for SAFSTOR • the estimated cost of decommissioning a BWR is lowest for OECON and highest for SAFSTOR. In all cases, SAFSTOR has the lowest occupational radiation dose and the highest cost.

  14. http://arXiv.org/physics/0507088 Teaching About Nature's Nuclear Reactors

    E-Print Network [OSTI]

    Learned, John

    http://arXiv.org/physics/0507088 Teaching About Nature's Nuclear Reactors J. Marvin Herndon reactors existed in uranium deposits on Earth long before Enrico Fermi built the first man-made nuclear reactors. The subject of planetocentric nuclear fission reactors can be a jumping off point for stimulating

  15. Nuclear Reactor Safeguards and Monitoring with Antineutrino Detectors

    E-Print Network [OSTI]

    Adam Bernstein; Yifang Wang; Giorgio Gratta; Todd West

    2001-08-01T23:59:59.000Z

    Cubic-meter-sized antineutrino detectors can be used to non-intrusively, robustly and automatically monitor and safeguard a wide variety of nuclear reactor types, including power reactors, research reactors, and plutonium production reactors. Since the antineutrino spectra and relative yields of fissioning isotopes depend on the isotopic composition of the core, changes in composition can be observed without ever directly accessing the core itself. Information from a modest-sized antineutrino detector, coupled with the well-understood principles that govern the core's evolution in time, can be used to determine whether the reactor is being operated in an illegitimate way. A group at Sandia is currently constructing a one cubic meter antineutrino detector at the San Onofre reactor site in California to demonstrate these principles.

  16. Spent nuclear fuel discharges from US reactors 1993

    SciTech Connect (OSTI)

    Not Available

    1995-02-01T23:59:59.000Z

    The Energy Information Administration (EIA) of the U.S. Department of Energy (DOE) administers the Nuclear Fuel Data Survey, Form RW-859. This form is used to collect data on fuel assemblies irradiated at commercial nuclear reactors operating in the United States, and the current inventories and storage capacities of those reactors. These data are important to the design and operation of the equipment and facilities that DOE will use for the future acceptance, transportation, and disposal of spent fuels. The data collected and presented identifies trends in burnup, enrichment, and spent nuclear fuel discharged form commercial light-water reactor as of December 31, 1993. The document covers not only spent nuclear fuel discharges; but also site capacities and inventories; canisters and nonfuel components; and assembly type characteristics.

  17. Magnitude and reactivity consequences of moisture ingress into the modular High-Temperature Gas-Cooled Reactor core

    SciTech Connect (OSTI)

    Smith, O.L. (Oak Ridge National Lab., TN (United States))

    1992-12-01T23:59:59.000Z

    Inadvertent admission of moisture into the primary system of a modular high-temperature gas-cooled reactor has been identified in US Department of Energy-sponsored studies as an important safety concern. The work described here develops an analytical methodology to quantify the pressure and reactivity consequences of steam-generator tube rupture and other moisture-ingress-related incidents. Important neutronic and thermohydraulic processes are coupled with reactivity feedback and safety and control system responses. The rate and magnitude of steam buildup are found to be dominated by major system features such as break size compared with safety valve capacity and reliability and less sensitive to factors such as heat transfer coefficients. The results indicate that ingress transients progress at a slower pace than previously predicted by bounding analyses, with milder power overshoots and more time for operator or automatic corrective actions.

  18. REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

    SciTech Connect (OSTI)

    Nichols, T.; Beals, D.; Sternat, M.

    2011-07-18T23:59:59.000Z

    Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Many research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical analysis. The fuel assembly was modeled using MONTEBURNS(MCNP5/ ORIGEN2.2) and MCNPX/CINDER90. The results from the models have been compared to each other and to the measured data.

  19. Distributed expert systems for nuclear reactor control

    SciTech Connect (OSTI)

    Otaduy, P.J.

    1992-12-01T23:59:59.000Z

    A network of distributed expert systems is the heart of a prototype supervisory control architecture developed at the Oak Ridge National Laboratory (ORNL) for an advanced multimodular reactor. Eight expert systems encode knowledge on signal acquisition, diagnostics, safeguards, and control strategies in a hybrid rule-based, multiprocessing and object-oriented distributed computing environment. An interactive simulation of a power block consisting of three reactors and one turbine provides a realistic, testbed for performance analysis of the integrated control system in real-time. Implementation details and representative reactor transients are discussed.

  20. Distributed expert systems for nuclear reactor control

    SciTech Connect (OSTI)

    Otaduy, P.J.

    1992-01-01T23:59:59.000Z

    A network of distributed expert systems is the heart of a prototype supervisory control architecture developed at the Oak Ridge National Laboratory (ORNL) for an advanced multimodular reactor. Eight expert systems encode knowledge on signal acquisition, diagnostics, safeguards, and control strategies in a hybrid rule-based, multiprocessing and object-oriented distributed computing environment. An interactive simulation of a power block consisting of three reactors and one turbine provides a realistic, testbed for performance analysis of the integrated control system in real-time. Implementation details and representative reactor transients are discussed.

  1. PHYSICS AND ENGINEERING OF NUCLEAR REACTORS AT THE ECOLE NATIONALE SUPRIEURE DE PHYSIQUE

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    PHYSICS AND ENGINEERING OF NUCLEAR REACTORS AT THE ECOLE NATIONALE SUPÃ?RIEURE DE PHYSIQUE DE. In this context, it is necessary to design and evaluate new generations of nuclear reactors as defined by the Gen. Introduction Nuclear reactors currently generate nearly one fourth of the electricity in the world, and nuclear

  2. Decision-support tool for assessing future nuclear reactor generation portfolios.

    E-Print Network [OSTI]

    Oosterlee, Cornelis W. "Kees"

    Decision-support tool for assessing future nuclear reactor generation portfolios. Shashi Jain, where especially capital costs are known to be highly uncertain. Differ- ent nuclear reactor types uncertainties in the cost elements of a nuclear power plant, to provide an optimal portfolio of nuclear reactors

  3. U.N. report concludes that Syrian site destroyed in 2007 was a nuclear reactor

    E-Print Network [OSTI]

    U.N. report concludes that Syrian site destroyed in 2007 was a nuclear reactor Joby Warrick, 24 May that Syria "very likely" was building a secret nuclear reactor in 2007 when the partially completed project, 2007, was a nuclear reactor intended for making fuel for nuclear bombs, a claim that Syria has

  4. Dual annular rotating "windowed" nuclear reflector reactor control system

    DOE Patents [OSTI]

    Jacox, Michael G. (Idaho Falls, ID); Drexler, Robert L. (Idaho Falls, ID); Hunt, Robert N. M. (Idaho Falls, ID); Lake, James A. (Idaho Falls, ID)

    1994-01-01T23:59:59.000Z

    A nuclear reactor control system is provided in a nuclear reactor having a core operating in the fast neutron energy spectrum where criticality control is achieved by neutron leakage. The control system includes dual annular, rotatable reflector rings. There are two reflector rings: an inner reflector ring and an outer reflector ring. The reflectors are concentrically assembled, surround the reactor core, and each reflector ring includes a plurality of openings. The openings in each ring are capable of being aligned or non-aligned with each other. Independent driving means for each of the annular reflector rings is provided so that reactor criticality can be initiated and controlled by rotation of either reflector ring such that the extent of alignment of the openings in each ring controls the reflection of neutrons from the core.

  5. 288 Int. J. Nuclear Energy Science and Technology, Vol. 7, No. 4, 2013 Multi-physics modelling of nuclear reactors

    E-Print Network [OSTI]

    Demazière, Christophe

    of nuclear reactors: current practices in a nutshell Christophe Demazière Department of Applied Physics of nuclear reactors are based on the use of different solvers for resolving the different physical fields and the corresponding approximations. Keywords: nuclear reactors; multi-physics; multi-scale; modelling; deterministic

  6. Raytheon explores thorium for next generation nuclear reactor

    SciTech Connect (OSTI)

    Crawford, M.

    1994-03-08T23:59:59.000Z

    Few new orders for nuclear power plants have been placed anywhere in the world in the last 20 years, but that is not discouraging Raytheon Engineers Constructors from making plans to explore new light water reactor technologies for commercial markets. The Lexington, Mass.-based company, which has extensive experience in nuclear power engineering and construction, has a vision for the light water reactor of the future - one that is based on the use of thorium-232, an element that decays over several steps to uranium-233. The use of thorium and a small amount of uranium that is 20 percent enriched is seen as providing operational, environmental, and safety advantages over reactors using the standard fuel mixture of uranium-238 and enriched uranium-235. According to Raytheon, the system could improve the economics of some reactors' operations by reducing fuel costs and lowering related waste volumes. At the same time, reactor safety could be improved by simpler control rod systems and the absence from reactor coolant of corrosive boric acid, which is used to slow neutrons in order to enhance reactions. Using thorium is also attractive because more of the fuel is burned up by the reactor, an estimated 12 percent as compared to about 4 percent for U-235. However, the technology's greatest attraction may well be its implications for nuclear proliferation. Growing plutonium inventories embedded in spent fuel rods from light water reactors have sparked concern worldwide. But according to Raytheon, using a thorium-based fuel core would alleviate this concern because it would produce only small quantities of plutonium. A thorium-based fuel system would produce 12 kilograms of plutonium over a decade versus 2,235 kilograms for an equivalent reactor operating with conventional uranium fuel.

  7. Cynod: A Neutronics Code for Pebble Bed Modular Reactor Coupled Transient Analysis

    SciTech Connect (OSTI)

    Hikaru Hiruta; Abderrafi M. Ougouag; Hans D. Gougar; Javier Ortensi

    2008-09-01T23:59:59.000Z

    The Pebble Bed Reactor (PBR) is one of the two concepts currently considered for development into the Next Generation Nuclear Plant (NGNP). This interest is due, in particular, to the concept’s inherent safety characteristics. In order to verify and confirm the design safety characteristics of the PBR computational tools must be developed that treat the range of phenomena that are expected to be important for this type of reactors. This paper presents a recently developed 2D R-Z cylindrical nodal kinetics code and shows some of its capabilities by applying it to a set of known and relevant benchmarks. The new code has been coupled to the thermal hydraulics code THERMIX/KONVEK[1] for application to the simulation of very fast transients in PBRs. The new code, CYNOD, has been written starting with a fixed source solver extracted from the nodal cylindrical geometry solver contained within the PEBBED code. The fixed source solver was then incorporated into a kinetic solver.. The new code inherits the spatial solver characteristics of the nodal solver within PEBBED. Thus, the time-dependent neutron diffusion equation expressed analytically in each node of the R-Z cylindrical geometry sub-domain (or node) is transformed into one-dimensional equations by means of the usual transverse integration procedure. The one-dimensional diffusion equations in each of the directions are then solved using the analytic Green’s function method. The resulting equations for the entire domain are then re-cast in the form of the Direct Coarse Mesh Finite Difference (D-CMFD) for convenience of solution. The implicit Euler method is used for the time variable discretization. In order to correctly treat the cusping effect for nodes that contain a partially inserted control rod a method is used that takes advantage of the Green’s function solution available in the intrinsic method. In this corrected treatment, the nodes are re-homogenized using axial flux shapes reconstructed based on the Green’s function method. The performance of the new code is demonstrated by applying it to a delayed supercritical problem and a to the OECD PBMR400 rod ejection benchmark problem. The latter makes use of the coupled CYNOD-THERMIX/KONVEK codes. A final improvement to the code is the subject of a companion paper: a heterogeneous TRISO fuel particle model was devised and incorporated into the code and used to provide an enhanced Doppler treatment. The new code is currently being coupled to the RELAP5-3D code for thermal-hydraulics. The full length paper will include extensive summaries of the equations and algorithm, descriptions of the sample and benchmark problems and details of the results. It is shown, in inter-code comparisons, that the new code correctly predicts the transient behaviors of the test problems.

  8. Spectral Structure of Electron Antineutrinos from Nuclear Reactors

    E-Print Network [OSTI]

    D. A. Dwyer; T. J. Langford

    2014-07-04T23:59:59.000Z

    Recent measurements of the positron energy spectrum obtained from inverse beta decay interactions of reactor electron antineutrinos show an excess in the 4 to 6 MeV region relative to current predictions. First-principle calculations of fission and beta decay processes within a typical pressurized water reactor core identify prominent fission daughter isotopes as a possible origin for this excess. These calculations also predict percent-level substructure in the antineutrino spectrum due to Coulomb effects in beta decay. Precise measurement of this substructure can constrain nuclear reactor physics. The substructure can be a systematic uncertainty for measurements utilizing the detailed spectral shape.

  9. Spectral Structure of Electron Antineutrinos from Nuclear Reactors

    E-Print Network [OSTI]

    Dwyer, D A

    2014-01-01T23:59:59.000Z

    Recent measurements of the positron energy spectrum obtained from inverse beta decay interactions of reactor electron antineutrinos show an excess in the 4 to 6 MeV region relative to current predictions. First-principle calculations of fission and beta decay processes within a typical pressurized water reactor core identify prominent fission daughter isotopes as a possible origin for this excess. These calculations also predict percent-level substructure in the antineutrino spectrum due to Coulomb effects in beta decay. Precise measurement of this substructure can constrain nuclear reactor physics. The substructure can be a systematic uncertainty for measurements utilizing the detailed spectral shape.

  10. Assessment of passive decay heat removal in the General Atomics Modular Helium Reactor 

    E-Print Network [OSTI]

    Cocheme, Francois Guilhem

    2005-02-17T23:59:59.000Z

    that began in the 1950?s. The concept features a helium gas cooled reactor with a graphite moderated prismatic core that contains TRISO fuel. This study evaluates the passive decay heat removal capabilities of the MHR under abnormal conditions, more...

  11. University Reactor Conversion Lessons Learned Workshop for Texas A&M University Nuclear Science Center Reactor

    SciTech Connect (OSTI)

    Eric C. Woolstenhulme; Dana M. Meyer

    2007-04-01T23:59:59.000Z

    The objectives of this meeting were to capture the observations, insights, issues, concerns, and ideas of those involved in the Texas A&M University Nuclear Science Center (TAMU NSC) TRIGA Reactor Conversion so that future efforts can be conducted with greater effectiveness, efficiency, and with fewer challenges. This workshop was held in conjunction with a similar workshop for the University of Florida Reactor Conversion. Some of the generic lessons from that workshop are included in this report for completeness.

  12. Low exchange element for nuclear reactor

    DOE Patents [OSTI]

    Brogli, Rudolf H. (Aarau, CH); Shamasunder, Bangalore I. (Encinitas, CA); Seth, Shivaji S. (Encinitas, CA)

    1985-01-01T23:59:59.000Z

    A flow exchange element is presented which lowers temperature gradients in fuel elements and reduces maximum local temperature within high temperature gas-cooled reactors. The flow exchange element is inserted within a column of fuel elements where it serves to redirect coolant flow. Coolant which has been flowing in a hotter region of the column is redirected to a cooler region, and coolant which has been flowing in the cooler region of the column is redirected to the hotter region. The safety, efficiency, and longevity of the high temperature gas-cooled reactor is thereby enhanced.

  13. Production capabilities in US nuclear reactors for medical radioisotopes

    SciTech Connect (OSTI)

    Mirzadeh, S.; Callahan, A.P.; Knapp, F.F. Jr. [Oak Ridge National Lab., TN (United States); Schenter, R.E. [Westinghouse Hanford Co., Richland, WA (United States)

    1992-11-01T23:59:59.000Z

    The availability of reactor-produced radioisotopes in the United States for use in medical research and nuclear medicine has traditionally depended on facilities which are an integral part of the US national laboratories and a few reactors at universities. One exception is the reactor in Sterling Forest, New York, originally operated as part of the Cintichem (Union Carbide) system, which is currently in the process of permanent shutdown. Since there are no industry-run reactors in the US, the national laboratories and universities thus play a critical role in providing reactor-produced radioisotopes for medical research and clinical use. The goal of this survey is to provide a comprehensive summary of these production capabilities. With the temporary shutdown of the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) in November 1986, the radioisotopes required for DOE-supported radionuclide generators were made available at the Brookhaven National Laboratory (BNL) High Flux Beam Reactor (HFBR). In March 1988, however, the HFBR was temporarily shut down which forced investigators to look at other reactors for production of the radioisotopes. During this period the Missouri University Research Reactor (MURR) played an important role in providing these services. The HFIR resumed routine operation in July 1990 at 85 MW power, and the HFBR resumed operation in June 1991, at 30 MW power. At the time of the HFBR shutdown, there was no available comprehensive overview which could provide information on status of the reactors operating in the US and their capabilities for radioisotope production. The obvious need for a useful overview was thus the impetus for preparing this survey, which would provide an up-to-date summary of those reactors available in the US at both the DOE-funded national laboratories and at US universities where service irradiations are currently or expected to be conducted.

  14. Safety Culture in the US Nuclear Regulatory Commission's Reactor Oversight Process

    Broader source: Energy.gov [DOE]

    Presenter: Undine Shoop, Chief, Health Physics and Human Performance Branch, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission

  15. Synergistic Smart Fuel For In-pile Nuclear Reactor Measurements

    SciTech Connect (OSTI)

    James A. Smith; Dale K. Kotter; Randall A. Ali; Steven L . Garrett

    2013-10-01T23:59:59.000Z

    In March 2011, an earthquake of magnitude 9.0 on the Richter scale struck Japan with its epicenter on the northeast coast, near the Tohoku region. In addition to the immense physical destruction and casualties across the country, several nuclear power plants (NPP) were affected. It was the Fukushima Daiichi NPP that experienced the most severe and irreversible damage. The earthquake brought the reactors at Fukushima to an automatic shutdown and because the power transmission lines were damaged, emergency diesel generators (EDGs) were activated to ensure that there was continued cooling of the reactors and spent fuel pools. The situation was being successfully managed until the tsunami hit about forty-five minutes later with a maximum wave height of approximately 15 m. The influx of water submerged the EDGs, the electrical switchgear, and dc batteries, resulting in the total loss of power to the reactors.2 At this point, the situation became critical. There was a loss of the sensors and instrumentation within the reactor that could have provided valuable information to guide the operators to make informed decisions and avoid the unfortunate events that followed. In the light of these events, we have developed and tested a potential self-powered thermoacoustic system, which will have the ability to serve as a temperature sensor and can transmit data independently of electronic networks. Such a device is synergistic with the harsh environment of the nuclear reactor as it utilizes the heat from the nuclear fuel to provide the input power.

  16. Foundational development of an advanced nuclear reactor integrated safety code.

    SciTech Connect (OSTI)

    Clarno, Kevin (Oak Ridge National Laboratory, Oak Ridge, TN); Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth (Ktech Corporation, Albuquerque, NM); Hooper, Russell Warren; Humphries, Larry LaRon

    2010-02-01T23:59:59.000Z

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  17. Identification of Selected Areas to Support Federal Clean Energy Goals Using Small Modular Reactors

    SciTech Connect (OSTI)

    Belles, Randy [ORNL; Mays, Gary T [ORNL; Omitaomu, Olufemi A [ORNL; Poore III, Willis P [ORNL

    2013-12-01T23:59:59.000Z

    This analysis identifies candidate locations, in a broad sense, where there are high concentrations of federal government agency use of electricity, which are also suitable areas for near-term SMRs. Near-term SMRs are based on light-water reactor (LWR) technology with compact design features that are expected to offer a host of safety, siting, construction, and economic benefits. These smaller plants are ideally suited for small electric grids and for locations that cannot support large reactors, thus providing utilities or governement entities with the flexibility to scale power production as demand changes by adding additional power by deploying more modules or reactors in phases. This research project is aimed at providing methodologies, information, and insights to assist the federal government in meeting federal clean energy goals.

  18. Automatic coolant flow control device for a nuclear reactor assembly

    DOE Patents [OSTI]

    Hutter, E.

    1984-01-27T23:59:59.000Z

    A device which controls coolant flow through a nuclear reactor assembly comprises a baffle means at the exit end of said assembly having a plurality of orifices, and a bimetallic member in operative relation to the baffle means such that at increased temperatures said bimetallic member deforms to unblock some of said orifices and allow increased coolant flow therethrough.

  19. advanced-gas-cooled-nuclear-reactor materials evaluation: Topics...

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

    advanced-gas-cooled-nuclear-reactor materials evaluation First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index...

  20. Passive cooling system for nuclear reactor containment structure

    DOE Patents [OSTI]

    Gou, Perng-Fei (Saratoga, CA); Wade, Gentry E. (Saratoga, CA)

    1989-01-01T23:59:59.000Z

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  1. Method of controlling crystallite size in nuclear-reactor fuels

    DOE Patents [OSTI]

    Lloyd, Milton H. (Oak Ridge, TN); Collins, Jack L. (Knoxville, TN); Shell, Sam E. (Oak Ridge, TN)

    1985-01-01T23:59:59.000Z

    Improved spherules for making enhanced forms of nuclear-reactor fuels are prepared by internal gelation procedures within a sol-gel operation and are accomplished by first boiling the concentrated HMTA-urea feed solution before engaging in the spherule-forming operation thereby effectively controlling crystallite size in the product spherules.

  2. Automatic coolant flow control device for a nuclear reactor assembly

    DOE Patents [OSTI]

    Hutter, Ernest (Wilmette, IL)

    1986-01-01T23:59:59.000Z

    A device which controls coolant flow through a nuclear reactor assembly comprises a baffle means at the exit end of said assembly having a plurality of orifices, and a bimetallic member in operative relation to the baffle means such that at increased temperatures said bimetallic member deforms to unblock some of said orifices and allow increased coolant flow therethrough.

  3. Natural circulating passive cooling system for nuclear reactor containment structure

    DOE Patents [OSTI]

    Gou, Perng-Fei (Saratoga, CA); Wade, Gentry E. (Saratoga, CA)

    1990-01-01T23:59:59.000Z

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  4. Safety program considerations for space nuclear reactor systems

    SciTech Connect (OSTI)

    Cropp, L.O.

    1984-08-01T23:59:59.000Z

    This report discusses the necessity for in-depth safety program planning for space nuclear reactor systems. The objectives of the safety program and a proposed task structure is presented for meeting those objectives. A proposed working relationship between the design and independent safety groups is suggested. Examples of safety-related design philosophies are given.

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

    E-Print Network [OSTI]

    Anitescu, Mihai

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

  6. Piccolo Micromegas: first in-core measurements in a nuclear reactor

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Piccolo Micromegas: first in-core measurements in a nuclear reactor J. Pancina , S. Andriamonjea in the coupling of an accelerator with a nuclear reactor. Such systems will need neutron detectors working domains. For the first time, Piccolo Micromegas has been placed in the core of a nuclear reactor

  7. FRP Retrofit of the Ring-Beam of a Nuclear Reactor Containment Structure

    E-Print Network [OSTI]

    SP·215-18 FRP Retrofit of the Ring-Beam of a Nuclear Reactor Containment Structure by M. Demers. A for the storage of the moderately contaminated nuclear reactor. The enforcement of more rigorous environmental. 1. HISTORY 1.1 Decommissioning of the Reactor The Gentilly-I nuclear power plant, located

  8. Status of Potential New Commercial Nuclear Reactors in the United Release Date: December 2007

    E-Print Network [OSTI]

    Noble, James S.

    1 Status of Potential New Commercial Nuclear Reactors in the United States Release Date: December for building new nuclear power reactors in the United States. Evidence of this includes press releases and conditionally operate new commercial nuclear reactors. Actual applications will also be included on future

  9. Identification and localization of absorbers of variable strength in nuclear reactors

    E-Print Network [OSTI]

    Demazière, Christophe

    Identification and localization of absorbers of variable strength in nuclear reactors C. Demazie evenly distrib- uted throughout the core of a commercial nuclear reactor. The novelty and ergodic in time, can be used for many diagnostic purposes in nuclear reactors. Many examples can be found

  10. Walking and Climbing Service Robots for Safety Inspection of Nuclear Reactor Pressure Vessels

    E-Print Network [OSTI]

    Chen, Sheng

    Walking and Climbing Service Robots for Safety Inspection of Nuclear Reactor Pressure Vessels B of Electronics and Computer Science, University of Southampton, Southampton, UK Abstract: Nuclear reactor and the usefulness of these robots for improving safety inspection of nuclear reactors in general are discussed

  11. Japanese set to direct `sun-power' nuclear reactor in France September 16, 2005

    E-Print Network [OSTI]

    Japanese set to direct `sun-power' nuclear reactor in France September 16, 2005 Japan has been develop three generations of nuclear reactors and includes six low-capacity experimental reactors and a 17 asked to nominate the chief of an international project to build a multi- billion-dollar nuclear fusion

  12. A NOVEL MICROMEGAS DETECTOR FOR IN-CORE NUCLEAR REACTOR NEUTRON FLUX MEASUREMENTS

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    1 A NOVEL MICROMEGAS DETECTOR FOR IN-CORE NUCLEAR REACTOR NEUTRON FLUX MEASUREMENTS S. ANDRIAMONJE Talence Cedex, France Future fast nuclear reactors designed for energy production and transmutation to neutron detection inside nuclear reactor is given. The advantage of this detector over conventional

  13. Operating strategy generators for nuclear reactors

    SciTech Connect (OSTI)

    Solovyev, D. A., E-mail: and@est.mephi.ru; Semenov, A. A.; Shchukin, N. V. [National Research Nuclear University MEPhI (Russian Federation)

    2011-12-15T23:59:59.000Z

    Operating strategy generators, i.e., the software intended for increasing the efficiency of work of nuclear power plant operators, are discussed. The possibilities provided by the domestic and foreign operating-strategy generators are analyzed.

  14. Nuclear Reactor Technologies | 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: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What's Possible for Renewable Energy:Nanowire3627 Federal Register /76 LosExperimentalSecurityReactor

  15. Design, Analysis and Optimization of the Power Conversion System for the Modular Pebble Bed Reactor System

    E-Print Network [OSTI]

    Pebble Bed Reactor system (MPBR) requires a gas turbine cycle (Brayton cycle) as the power conversion for the gas turbine cycle. The development of an initial reference design for an indirect helium cycle has for the system. Load transients simulations show that the indirect, three-shaft arrangement gas turbine power

  16. Breeding nuclear fuels with accelerators: replacement for breeder reactors

    SciTech Connect (OSTI)

    Grand, P.; Takahashi, H.

    1984-01-01T23:59:59.000Z

    One application of high energy particle accelerators has been, and still is, the production of nuclear fuel for the nuclear energy industry; tantalizing because it would create a whole new industry. This approach to producing fissile from fertile material was first considered in the early 1950's in the context of the nuclear weapons program. A considerable development effort was expended before discovery of uranium ore in New Mexico put an end to the project. Later, US commitment to the Liquid Metal Fast Breeder Reactors (LMFBR) killed any further interest in pursuing accelerator breeder technology. Interest in the application of accelerators to breed nuclear fuels, and possibly burn nuclear wastes, revived in the late 1970's, when the LMFBR came under attack during the Carter administration. This period gave the opportunity to revisit the concept in view of the present state of the technology. This evaluation and the extensive calculational modeling of target designs that have been carried out are promising. In fact, a nuclear fuel cycle of Light Water Reactors and Accelerator Breeders is competitive to that of the LMFBR. At this time, however, the relative abundance of uranium reserves vs electricity demand and projected growth rate render this study purely academic. It will be for the next generation of accelerator builders to demonstate the competitiveness of this technology versus that of other nuclear fuel cycles, such as LMFBR's or Fusion Hybrid systems. 22 references, 1 figure, 5 tables.

  17. Studies Related to the Oregon State University High Temperature Test Facility: Scaling, the Validation Matrix, and Similarities to the Modular High Temperature Gas-Cooled Reactor

    SciTech Connect (OSTI)

    Richard R. Schultz; Paul D. Bayless; Richard W. Johnson; William T. Taitano; James R. Wolf; Glenn E. McCreery

    2010-09-01T23:59:59.000Z

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5 year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant project. Because the NRC interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC). Since DOE has incorporated the HTTF as an ingredient in the NGNP thermal-fluids validation program, several important outcomes should be noted: 1. The reference prismatic reactor design, that serves as the basis for scaling the HTTF, became the modular high temperature gas-cooled reactor (MHTGR). The MHTGR has also been chosen as the reference design for all of the other NGNP thermal-fluid experiments. 2. The NGNP validation matrix is being planned using the same scaling strategy that has been implemented to design the HTTF, i.e., the hierarchical two-tiered scaling methodology developed by Zuber in 1991. Using this approach a preliminary validation matrix has been designed that integrates the HTTF experiments with the other experiments planned for the NGNP thermal-fluids verification and validation project. 3. Initial analyses showed that the inherent power capability of the OSU infrastructure, which only allowed a total operational facility power capability of 0.6 MW, is inadequate to permit steady-state operation at reasonable conditions. 4. To enable the HTTF to operate at a more representative steady-state conditions, DOE recently allocated funding via a DOE subcontract to HTTF to permit an OSU infrastructure upgrade such that 2.2 MW will become available for HTTF experiments. 5. Analyses have been performed to study the relationship between HTTF and MHTGR via the hierarchical two-tiered scaling methodology which has been used successfully in the past, e.g., APEX facility scaling to the Westinghouse AP600 plant. These analyses have focused on the relationship between key variables that will be measured in the HTTF to the counterpart variables in the MHTGR with a focus on natural circulation, using nitrogen as a working fluid, and core heat transfer. 6. Both RELAP5-3D and computational fluid dynamics (CD-Adapco’s STAR-CCM+) numerical models of the MHTGR and the HTTF have been constructed and analyses are underway to study the relationship between the reference reactor and the HTTF. The HTTF is presently being designed. It has ¼-scaling relationship to the MHTGR in both the height and the diameter. Decisions have been made to design the reactor cavity cooling system (RCCS) simulation as a boundary condition for the HTTF to ensure that (a) the boundary condition is well defined and (b) the boundary condition can be modified easily to achieve the desired heat transfer sink for HTTF experimental operations.

  18. Investigation of cracking and leaking of nuclear reactor pools

    E-Print Network [OSTI]

    Cooper, William Bernard

    1965-01-01T23:59:59.000Z

    *ion of the pool was lined with a 0. 2y in, aluminum I'xxer which hss since corroded. snd. pitted. allowing the dent nex slimed watel' used xxx the pool to come into dix'ec't contact with the concx etc pool wall. Pigure Hc, LacBtx one oi l&BkBg?e A mare detai3... good, Job i'or several years without, renovation, [0] This type of reactor has, however, developed. some maJor problems that have?or will, require r'enovation ~ Figure l. Nuclear Science Center Reactor Facie. ity at, Texas A& University...

  19. Heat barrier for use in a nuclear reactor facility

    DOE Patents [OSTI]

    Keegan, Charles P. (South Huntingdon Twp., Westmoreland County, PA)

    1988-01-01T23:59:59.000Z

    A thermal barrier for use in a nuclear reactor facility is disclosed herein. Generally, the thermal barrier comprises a flexible, heat-resistant web mounted over the annular space between the reactor vessel and the guard vessel in order to prevent convection currents generated in the nitrogen atmosphere in this space from entering the relatively cooler atmosphere of the reactor cavity which surrounds these vessels. Preferably, the flexible web includes a blanket of heat-insulating material formed from fibers of a refractory material, such as alumina and silica, sandwiched between a heat-resistant, metallic cloth made from stainless steel wire. In use, the web is mounted between the upper edges of the guard vessel and the flange of a sealing ring which surrounds the reactor vessel with a sufficient enough slack to avoid being pulled taut as a result of thermal differential expansion between the two vessels. The flexible web replaces the rigid and relatively complicated structures employed in the prior art for insulating the reactor cavity from the convection currents generated between the reactor vessel and the guard vessel.

  20. Synergistic smart fuel for in-pile nuclear reactor measurements

    SciTech Connect (OSTI)

    Smith, J.A.; Kotter, D.K. [Idaho National Laboratories, Idaho Falls (United States); Ali, R.A.; Garrett, S.L. [Penn State University, University Park, State College, PA 16801 (United States)

    2013-07-01T23:59:59.000Z

    The thermo-acoustic fuel rod sensor developed in this research has demonstrated a novel technique for monitoring the temperature within the core of a nuclear reactor or the temperature of the surrounding heat-transfer fluid. It uses the heat from the nuclear fuel to generate sustained acoustic oscillations whose frequency will be indicative of the temperature. Converting a nuclear fuel rod into this type of thermo-acoustic sensor simply requires the insertion of a porous material (stack). This sensor has demonstrated a synergy with the elevated temperatures that exist within the nuclear reactor using materials that have only minimal susceptibility to high-energy particle fluxes. When the sensor is in operation, the sound waves radiated from the fuel rod resonator will propagate through the surrounding cooling fluid. The frequency of these oscillations is directly correlated with an effective temperature within the fuel rod resonator. This device is self-powered and is operational even in case of total loss of power of the reactor.

  1. SNIF: A Futuristic Neutrino Probe for Undeclared Nuclear Fission Reactors

    E-Print Network [OSTI]

    Thierry Lasserre; Maximilien Fechner; Guillaume Mention; Romain Reboulleau; Michel Cribier; Alain Letourneau; David Lhuillier

    2010-11-16T23:59:59.000Z

    Today reactor neutrino experiments are at the cutting edge of fundamental research in particle physics. Understanding the neutrino is far from complete, but thanks to the impressive progress in this field over the last 15 years, a few research groups are seriously considering that neutrinos could be useful for society. The International Atomic Energy Agency (IAEA) works with its Member States to promote safe, secure and peaceful nuclear technologies. In a context of international tension and nuclear renaissance, neutrino detectors could help IAEA to enforce the Treaty on the Non-Proliferation of Nuclear Weapons (NPT). In this article we discuss a futuristic neutrino application to detect and localize an undeclared nuclear reactor from across borders. The SNIF (Secret Neutrino Interactions Finder) concept proposes to use a few hundred thousand tons neutrino detectors to unveil clandestine fission reactors. Beyond previous studies we provide estimates of all known background sources as a function of the detector's longitude, latitude and depth, and we discuss how they impact the detectability.

  2. Nuclear reactor spacer grid and ductless core component

    DOE Patents [OSTI]

    Christiansen, David W. (Kennewick, WA); Karnesky, Richard A. (Richland, WA)

    1989-01-01T23:59:59.000Z

    The invention relates to a nuclear reactor spacer grid member for use in a liquid cooled nuclear reactor and to a ductless core component employing a plurality of these spacer grid members. The spacer grid member is of the egg-shell type and is constructed so that the walls of the cell members of the grid member are formed of a single thickness of metal to avoid tolerance problems. Within each cell member is a hydraulic spring which laterally constrains the nuclear material bearing rod which passes through each cell member against a hardstop in response to coolant flow through the cell member. This hydraulic spring is also suitable for use in a water cooled nuclear reactor. A core component constructed of, among other components, a plurality of these spacer grid members, avoids the use of a full length duct by providing spacer sleeves about the sodium tubes passing through the spacer grid members at locations between the grid members, thereby maintaining a predetermined space between adjacent grid members.

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

    SciTech Connect (OSTI)

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

    2006-12-11T23:59:59.000Z

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

  4. ANALYSIS OF SEPCTRUM CHOICES FOR SMALL MODULAR REACTORS-PERFORMANCE AND DEVELOPMENT 

    E-Print Network [OSTI]

    Kafle, Nischal

    2011-04-26T23:59:59.000Z

    is not just limited reactor pressure vessel, but can also analyze 1D, 2D, and 3D array of volumes and internal junctions ( The RELAP5-3D Code Development Team, 2005). RELAP5 contains three major level structures as shown in Fig 4; first is the input... equations such as conversation of mass momentum and energy and other field equations for pressure, velocity, specific internal energy, void fraction. RELAP5 code is one of the software for extensive thermal hydraulics and accident analysis. The brief...

  5. The passive safety characteristics of modular high temperature gas-cooled reactor fuel elements

    SciTech Connect (OSTI)

    Goodin, D.T.; Kania, M.J.; Nabielek, H.; Schenk, W.; Verfondern, K.

    1988-01-01T23:59:59.000Z

    High-Temperature Gas-Cooled Reactors (HTGR) in both the US and West Germany use an all-ceramic, coated fuel particle to retain fission products. Data from irradiation, postirradiation examinations and postirradiation heating experiments are used to study the performance capabilities of the fuel particles. The experimental results from fission product release tests with HTGR fuel are discussed. These data are used for development of predictive fuel performance models for purposes of design, licensing, and risk analyses. During off normal events, where temperatures may reach up to 1600/degree/C, the data show that no significant radionuclide releases from the fuel will occur.

  6. Assessment of passive decay heat removal in the General Atomics Modular Helium Reactor

    E-Print Network [OSTI]

    Cocheme, Francois Guilhem

    2005-02-17T23:59:59.000Z

    incentive is to separate the two plants enough for the hydrogen plant to be considered outside of the nuclear plant influence. The radioactivity level on the chemical side must be sufficiently low to avoid classifying the hydrogen production plant as a...

  7. Deployment at the Savannah River Site of a standardized, modular transportable and connectable hazard category 2 nuclear system for repackaging TRU waste

    SciTech Connect (OSTI)

    Lussiez, G. (Guy); Hickman, S. (Scott); Anast, K. R. (Kurt R.); Oliver, W. B. (William B.)

    2004-01-01T23:59:59.000Z

    This paper describes the conception, design, fabrication and deployment of a modular, transportable, connectable Category 2 nuclear system deployed at the Savannah River site to be used for characterizing and repackaging Transuranic Waste destined for the Waste Isolation Pilot Plant (WIPP). A standardized Nuclear Category 2 and Performance Category 2 envelope called a 'Nuclear Transportainer' was conceived and designed that provides a safety envelope for nuclear operations. The Nuclear Transportainer can be outfitted with equipment that performs functions necessary to meet mission objectives, in this case repackaging waste for shipment to WIPP. Once outfitted with process and ventilation systems the Nuclear Transportainer is a Modular Unit (MU). Each MU is connectable to other MUS - nuclear or non-nuclear - allowing for multiple functions, command & control, or increasing capacity. The design took advantage of work already in-progress at Los Alamos National Laboratory (LANL) for a similar system to be deployed at LANL's Technical Area 54.

  8. Passive heat-transfer means for nuclear reactors. [LMFBR

    DOE Patents [OSTI]

    Burelbach, J.P.

    1982-06-10T23:59:59.000Z

    An improved passive cooling arrangement is disclosed for maintaining adjacent or related components of a nuclear reactor within specified temperature differences. Specifically, heat pipes are operatively interposed between the components, with the vaporizing section of the heat pipe proximate the hot component operable to cool it and the primary condensing section of the heat pipe proximate the other and cooler component operable to heat it. Each heat pipe further has a secondary condensing section that is located outwardly beyond the reactor confinement and in a secondary heat sink, such as air ambient the containment, that is cooler than the other reactor component. By having many such heat pipes, an emergency passive cooling system is defined that is operative without electrical power.

  9. COMSOL-based Nuclear Reactor Kinetics Studies at the HFIR

    SciTech Connect (OSTI)

    Chandler, David [ORNL] [ORNL; Freels, James D [ORNL] [ORNL; Maldonado, G Ivan [ORNL] [ORNL; Primm, Trent [ORNL] [ORNL

    2011-01-01T23:59:59.000Z

    The computational ability to accurately predict the dynamic behavior of a nuclear reactor core in response to reactivity-induced perturbations is an important subject in reactor physics. Space-time and point kinetics methodologies were developed for the purpose of studying the transient-induced behavior of the High Flux Isotope Reactor s (HFIR) compact core. The space-time simulations employed the three-energy-group neutron diffusion equations, and transients initiated by control cylinder and hydraulic tube rabbit ejections were studied. The work presented here is the first step towards creating a comprehensive multiphysics methodology for studying the dynamic behavior of the HFIR core during reactivity perturbations. The results of these studies show that point kinetics is adequate for small perturbations in which the power distribution is assumed to be time-independent, but space-time methods must be utilized to determine localized effects.

  10. Support arrangements for core modules of nuclear reactors. [PWR

    DOE Patents [OSTI]

    Bollinger, L.R.

    1983-11-03T23:59:59.000Z

    A support arrangement is provided for the core modules of a nuclear reactor which provides support access through the control drive mechanisms of the reactor. This arrangement provides axial support of individual reactor core modules from the pressure vessel head in a manner which permits attachment and detachment of the modules from the head to be accomplished through the control drive mechanisms after their leadscrews have been removed. The arrangement includes a module support nut which is suspended from the pressure vessel head and screw threaded to the shroud housing for the module. A spline lock prevents loosening of the screw connection. An installation tool assembly, including a cell lifting and preloading tool and a torquing tool, fits through the control drive mechanism and provides lifting of the shroud housing while disconnecting the spline lock, as well as application of torque to the module support nut.

  11. Collecting and recirculating condensate in a nuclear reactor containment

    DOE Patents [OSTI]

    Schultz, Terry L. (Murrysville Boro, PA)

    1993-01-01T23:59:59.000Z

    An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank.

  12. Collecting and recirculating condensate in a nuclear reactor containment

    DOE Patents [OSTI]

    Schultz, T.L.

    1993-10-19T23:59:59.000Z

    An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank. 3 figures.

  13. Removable check valve for use in a nuclear reactor

    DOE Patents [OSTI]

    Dunn, Charlton (Calabasas, CA); Gutzmann, Edward A. (Simi Valley, CA)

    1988-01-01T23:59:59.000Z

    A removable check valve for interconnecting the discharge duct of a pump and an inlet coolant duct of a reactor core in a pool-type nuclear reactor. A manifold assembly is provided having an outer periphery affixed to and in fluid communication with the discharge duct of the pump and has an inner periphery having at least one opening therethrough. A housing containing a check valve is located within the inner periphery of the manifold. The upper end of the housing has an opening in alignment with the opening in the manifold assembly, and seals are provided above and below the openings. The lower end of the housing is adapted for fluid communication with the inlet duct of the reactor core.

  14. Software reliability and safety in nuclear reactor protection systems

    SciTech Connect (OSTI)

    Lawrence, J.D. [Lawrence Livermore National Lab., CA (United States)

    1993-11-01T23:59:59.000Z

    Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor.

  15. Advanced Nuclear Reactors | 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: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office511041cloth DocumentationProducts (VAP) VAP7-0973 1BP-14 PowerAdvanced Modeling &Advanced Nuclear

  16. Systems and methods for dismantling a nuclear reactor

    DOE Patents [OSTI]

    Heim, Robert R; Adams, Scott Ryan; Cole, Matthew Denver; Kirby, William E; Linnebur, Paul Damon

    2014-10-28T23:59:59.000Z

    Systems and methods for dismantling a nuclear reactor are described. In one aspect the system includes a remotely controlled heavy manipulator ("manipulator") operatively coupled to a support structure, and a control station in a non-contaminated portion of a workspace. The support structure provides the manipulator with top down access into a bioshield of a nuclear reactor. At least one computing device in the control station provides remote control to perform operations including: (a) dismantling, using the manipulator, a graphite moderator, concrete walls, and a ceiling of the bioshield, the manipulator being provided with automated access to all internal portions of the bioshield; (b) loading, using the manipulator, contaminated graphite blocks from the graphite core and other components from the bioshield into one or more waste containers; and (c) dispersing, using the manipulator, dust suppression and contamination fixing spray to contaminated matter.

  17. A comparison of nuclear reactor control room display panels

    E-Print Network [OSTI]

    Bowers, Frances Renae

    1988-01-01T23:59:59.000Z

    energized (and operating normally) and green indicates safe or deenergized (operation stopped) . This second coding system is the primary code in the control room of the TAMU Nuclear Engineering 10 Department's Research Reactor. It originated from... the red for danger standard mentioned at the beginning of this paragraph. In electrical systems, a closed circuit means that it is energized and therefore dangerous to touch. An open circuit means that current is not flowing and it is therefore safe...

  18. Search for Neutrino Oscillations at the Palo Verde Nuclear Reactors

    E-Print Network [OSTI]

    F. Boehm; J. Busenitz; B. Cook; G. Gratta; H. Henrikson; J. Kornis; D. Lawrence; K. B. Lee; K. McKinny; L. Miller; V. Novikov; A. Piepke; B. Ritchie; D. Tracy; P. Vogel; Y-F. Wang; J. Wolf

    1999-12-22T23:59:59.000Z

    We report on the initial results from a measurement of the anti-neutrino flux and spectrum at a distance of about 800 m from the three reactors of the Palo Verde Nuclear Generating Station using a segmented gadolinium-loaded scintillation detector. We find that the anti-neutrino flux agrees with that predicted in the absence of oscillations to better than 5%, excluding at 90% CL $\\rm\\bar\

  19. Expert system for online surveillance of nuclear reactor coolant pumps

    DOE Patents [OSTI]

    Gross, Kenny C. (Bolingbrook, IL); Singer, Ralph M. (Naperville, IL); Humenik, Keith E. (Columbia, MD)

    1993-01-01T23:59:59.000Z

    An expert system for online surveillance of nuclear reactor coolant pumps. This system provides a means for early detection of pump or sensor degradation. Degradation is determined through the use of a statistical analysis technique, sequential probability ratio test, applied to information from several sensors which are responsive to differing physical parameters. The results of sequential testing of the data provide the operator with an early warning of possible sensor or pump failure.

  20. Detachable connection for a nuclear reactor fuel assembly

    DOE Patents [OSTI]

    Christiansen, D.W.; Karnesky, R.A.

    1983-08-29T23:59:59.000Z

    A locking connection for releasably attaching a handling socket to the duct tube of a fuel assembly for a nuclear reactor. The connection comprises a load pad housing mechanically attached to the duct tube and a handling socket threadably secured within the housing. A retaining ring is interposed between the housing and the handling socket and is formed with a projection and depression engagable within a cavity and groove of the housing and handling socket, respectively, to form a detachable interlocked connection assembly.

  1. A comparison of nuclear reactor control room display panels 

    E-Print Network [OSTI]

    Bowers, Frances Renae

    1988-01-01T23:59:59.000Z

    . Evaluation Procedure 14 15 16 17 28 Analysis. 35 III RESULTS. 37 Discussion. 43 IV CONCLUSIONS AND RECOMMENDATIONS Recommendations REFERENCES 50 51 54 APPENDIX A: REACTOR OPERATOR INFORMATION PACKAGE. . . . . 57 APPENDIX B: AGN-201 STANDARD... related to human factors included poor training, inaccessibility or unavailability of pertinent information, copious amounts of unusable information and obscured displays. Since the Three Mile Island (TMI) accident, the U. S. Nuclear Regulatory...

  2. INVESTIGATIONS ON NUCLEAR SPECTROSCOPY AT THE REACTOR AND THEIR APPLICATIONS1

    E-Print Network [OSTI]

    Titov, Anatoly

    1 INVESTIGATIONS ON NUCLEAR SPECTROSCOPY AT THE REACTOR AND THEIR APPLICATIONS1 I.A. Kondurov , E. However the first work on nuclear spectroscopy was carried out before the reactor was launched; namely.M. Korotkikh, Yu.E. Loginov, V.V. Martynov Introduction Physical launch of the WWR-M reactor in the branch

  3. Granular flow in pebble-bed nuclear reactors: Scaling, Dust Generation, and Stress

    E-Print Network [OSTI]

    Rycroft, Chris H.

    Granular flow in pebble-bed nuclear reactors: Scaling, Dust Generation, and Stress Chris H. Keywords: granular flow, dust generation, numerical methods 1. Introduction Pebble-bed nuclear reactors prototypes of pebble-bed reactors, significant quantities of graphite dust have been observed due to rubbing

  4. Small Modular Reactors, National Security and Clean Energy: A U.S. National

    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: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of ScienceandMesa del Sol HomeFacebookScholarship Fund3Biology| National Nuclear SecurityDepartment

  5. Method for passive cooling liquid metal cooled nuclear reactors, and system thereof

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Busboom, Herbert J. (San Jose, CA)

    1991-01-01T23:59:59.000Z

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel.

  6. Research in nondestructive evaluation techniques for nuclear reactor concrete structures

    SciTech Connect (OSTI)

    Clayton, Dwight; Smith, Cyrus [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States)

    2014-02-18T23:59:59.000Z

    The purpose of the Materials Aging and Degradation (MAaD) Pathway of the Department of Energy's Light Water Reactor Sustainability (LWRS) Program is to develop the scientific basis for understanding and predicting longterm environmental degradation behavior of material in nuclear power plants and to provide data and methods to assess the performance of systems, structures, and components (SSCs) essential to safe and sustained nuclear power plant operations. The understanding of aging-related phenomena and their impacts on SSCs is expected to be a significant issue for any nuclear power plant planning for long-term operations (i.e. service beyond the initial license renewal period). Management of those phenomena and their impacts during long-term operations can be better enable by improved methods and techniques for detection, monitoring, and prediction of SSC degradation. The MAaD Pathway R and D Roadmap for Concrete, 'Light Water Reactor Sustainability Nondestructive Evaluation for Concrete Research and Development Roadmap', focused initial research efforts on understanding the recent concrete issues at nuclear power plants and identifying the availability of concrete samples for NDE techniques evaluation and testing. [1] An overview of the research performed by ORNL in these two areas is presented here.

  7. Technical Needs for Prototypic Prognostic Technique Demonstration for Advanced Small Modular Reactor Passive Components

    SciTech Connect (OSTI)

    Meyer, Ryan M.; Coble, Jamie B.; Hirt, Evelyn H.; Ramuhalli, Pradeep; Mitchell, Mark R.; Wootan, David W.; Berglin, Eric J.; Bond, Leonard J.; Henager, Charles H.

    2013-05-17T23:59:59.000Z

    This report identifies a number of requirements for prognostics health management of passive systems in AdvSMRs, documents technical gaps in establishing a prototypical prognostic methodology for this purpose, and describes a preliminary research plan for addressing these technical gaps. AdvSMRs span multiple concepts; therefore a technology- and design-neutral approach is taken, with the focus being on characteristics that are likely to be common to all or several AdvSMR concepts. An evaluation of available literature is used to identify proposed concepts for AdvSMRs along with likely operational characteristics. Available operating experience of advanced reactors is used in identifying passive components that may be subject to degradation, materials likely to be used for these components, and potential modes of degradation of these components. This information helps in assessing measurement needs for PHM systems, as well as defining functional requirements of PHM systems. An assessment of current state-of-the-art approaches to measurements, sensors and instrumentation, diagnostics and prognostics is also documented. This state-of-the-art evaluation, combined with the requirements, may be used to identify technical gaps and research needs in the development, evaluation, and deployment of PHM systems for AdvSMRs. A preliminary research plan to address high-priority research needs for the deployment of PHM systems to AdvSMRs is described, with the objective being the demonstration of prototypic prognostics technology for passive components in AdvSMRs. Greater efficiency in achieving this objective can be gained through judicious selection of materials and degradation modes that are relevant to proposed AdvSMR concepts, and for which significant knowledge already exists. These selections were made based on multiple constraints including the analysis performed in this document, ready access to laboratory-scale facilities for materials testing and measurement, and potential synergies with other national laboratory and university partners.

  8. RIS-M-2575 REFERENCE NEUTRON RADIOGRAPHS OF NUCLEAR REACTOR FUEL

    E-Print Network [OSTI]

    RISØ-M-2575 REFERENCE NEUTRON RADIOGRAPHS OF NUCLEAR REACTOR FUEL J. C. Domanus Abstract. Reference neutron radiographs of nuclear reactor fuel were produced by the Euraton Neutron Radiography Working Group-conpacted nuclear fuel pins. Those parts of the pins are shown where changes of ap- pearance differ from those

  9. Nuclear reactor power for an electrically powered orbital transfer vehicle

    SciTech Connect (OSTI)

    Jaffe, L.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Fujita, T.; Grossman, M.; Kia, T.; Nesmith, B.

    1987-01-01T23:59:59.000Z

    To help determine the systems requirements for a 300-kWe space nuclear reactor power system, a mission and spacecraft have been examined which utilize electric propulsion and this nuclear reactor power for multiple transfers of cargo between low Earth orbit (LEO) and geosynchronous Earth orbit (GEO). A propulsion system employing ion thrusters and xenon propellant was selected. Propellant and thrusters are replaced after each sortie to GEO. The mass of the Orbital Transfer Vehicle (OTV), empty and dry, is 11,000 kg; nominal propellant load is 5000 kg. The OTV operates between a circular orbit at 925 km altitude, 28.5 deg inclination, and GEO. Cargo is brought to the OTV by Shuttle and an Orbital Maneuvering Vehicle (OMV); the OTV then takes it to GEO. The OTV can also bring cargo back from GEO, for transfer by OMV to the Shuttle. OTV propellant is resupplied and the ion thrusters are replaced by the OMV before each trip to GEO. At the end of mission life, the OTV's electric propulsion is used to place it in a heliocentric orbit so that the reactor will not return to Earth. The nominal cargo capability to GEO is 6000 kg with a transit time of 120 days; 1350 kg can be transferred in 90 days, and 14,300 kg in 240 days. These capabilities can be considerably increased by using separate Shuttle launches to bring up propellant and cargo, or by changing to mercury propellant.

  10. Reactor Vessel and Reactor Vessel Internals Segmentation at Zion Nuclear Power Station - 13230

    SciTech Connect (OSTI)

    Cooke, Conrad; Spann, Holger [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)] [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)

    2013-07-01T23:59:59.000Z

    Zion Nuclear Power Station (ZNPS) is a dual-unit Pressurized Water Reactor (PWR) nuclear power plant located on the Lake Michigan shoreline, in the city of Zion, Illinois approximately 64 km (40 miles) north of Chicago, Illinois and 67 km (42 miles) south of Milwaukee, Wisconsin. Each PWR is of the Westinghouse design and had a generation capacity of 1040 MW. Exelon Corporation operated both reactors with the first unit starting production of power in 1973 and the second unit coming on line in 1974. The operation of both reactors ceased in 1996/1997. In 2010 the Nuclear Regulatory Commission approved the transfer of Exelon Corporation's license to ZionSolutions, the Long Term Stewardship subsidiary of EnergySolutions responsible for the decommissioning of ZNPS. In October 2010, ZionSolutions awarded Siempelkamp Nuclear Services, Inc. (SNS) the contract to plan, segment, remove, and package both reactor vessels and their respective internals. This presentation discusses the tools employed by SNS to remove and segment the Reactor Vessel Internals (RVI) and Reactor Vessels (RV) and conveys the recent progress. SNS's mechanical segmentation tooling includes the C-HORCE (Circumferential Hydraulically Operated Cutting Equipment), BMT (Bolt Milling Tool), FaST (Former Attachment Severing Tool) and the VRS (Volume Reduction Station). Thermal segmentation of the reactor vessels will be accomplished using an Oxygen- Propane cutting system. The tools for internals segmentation were designed by SNS using their experience from other successful reactor and large component decommissioning and demolition (D and D) projects in the US. All of the designs allow for the mechanical segmentation of the internals remotely in the water-filled reactor cavities. The C-HORCE is designed to saw seven circumferential cuts through the Core Barrel and Thermal Shield walls with individual thicknesses up to 100 mm (4 inches). The BMT is designed to remove the bolts that fasten the Baffle Plates to the Baffle Former Plates. The FaST is designed to remove the Baffle Former Plates from the Core Barrel. The VRS further volume reduces segmented components using multiple configurations of the 38i and horizontal reciprocating saws. After the successful removal and volume reduction of the Internals, the RV will be segmented using a 'First in the US' thermal cutting process through a co-operative effort with Siempelkamp NIS Ingenieurgesellschaft mbH using their experience at the Stade NPP and Karlsruhe in Germany. SNS mobilized in the fall of 2011 to commence execution of the project in order to complete the RVI segmentation, removal and packaging activities for the first unit (Unit 2) by end of the 2012/beginning 2013 and then mobilize to the second unit, Unit 1. Parallel to the completion of the segmentation of the reactor vessel internals at Unit 1, SNS will segment the Unit 2 pressure vessel and at completion move to Unit 1. (authors)

  11. Monitoring system for a liquid-cooled nuclear fission reactor

    DOE Patents [OSTI]

    DeVolpi, Alexander (Bolingbrook, IL)

    1987-01-01T23:59:59.000Z

    A monitoring system for detecting changes in the liquid levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting changes in the density of the liquid in these regions. A plurality of gamma radiation detectors are used, arranged vertically along the outside of the reactor vessel, and collimator means for each detector limits the gamma-radiation it receives as emitting from only isolated regions of the vessel. Excess neutrons produced by the fission reaction will be captured by the water coolant, by the steel reactor walls, or by the fuel or control structures in the vessel. Neutron capture by steel generates gamma radiation having an energy level of the order of 5-12 MeV, whereas neutron capture by water provides an energy level of approximately 2.2 MeV, and neutron capture by the fission fuel or its cladding provides an energy level of 1 MeV or less. The intensity of neutron capture thus changes significantly at any water-metal interface. Comparative analysis of adjacent gamma detectors senses changes from the normal condition with liquid coolant present to advise of changes in the presence and/or density of the coolant at these specific regions. The gamma detectors can also sense fission-product gas accumulation at the reactor head to advise of a failure of fuel-pin cladding.

  12. Commercial nuclear reactors and waste: the current status

    SciTech Connect (OSTI)

    Platt, A.M.; Robinson, J.V.

    1980-04-01T23:59:59.000Z

    During the last five years, the declared size of the commercial light water reactor (LWR) nuclear power industry in the US has steadily decreased. As of January 1980, the total number of power plants had dropped to 191 from the 226 in December 31, 1974. At least another nine were cancelled in the last few months. This report was developed as the first of a series to track implications to waste management due to such changes in the declared size of the industry. For the presently declared size, key conclusions are: the declared reactors will peak at a capacity of 162 GWe and consume about 10/sup 6/ MTU as enrichment feed. As few as two repositories of about 100,000 MTHM capacity each would hold the waste. Predisposal storage (reactor basins and AFRs) would peak at less than 100,000 MTHM (in the year 2020) with one repository opening in the year 1997 and the other in the year 2020. Most of the 100,000 MTHM would have to be in AFR storage unless current practice regarding reactor basin size was radically changed.

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

    DOE Patents [OSTI]

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

    1990-01-01T23:59:59.000Z

    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.

  14. Summary of space nuclear reactor power systems, 1983--1992

    SciTech Connect (OSTI)

    Buden, D.

    1993-08-11T23:59:59.000Z

    This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

  15. 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.

    2010-01-01T23:59:59.000Z

    Charges Relating to Nuclear Reactor Safety," 1976, availableissues impor tant to nuclear reactor safety. This report wasstudies of overall nuclear reactor safety have been

  16. 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.

    2010-01-01T23:59:59.000Z

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

  17. Identifying and bounding uncertainties in nuclear reactor thermal power calculations

    SciTech Connect (OSTI)

    Phillips, J.; Hauser, E.; Estrada, H. [Cameron, 1000 McClaren Woods Drive, Coraopolis, PA 15108 (United States)

    2012-07-01T23:59:59.000Z

    Determination of the thermal power generated in the reactor core of a nuclear power plant is a critical element in the safe and economic operation of the plant. Direct measurement of the reactor core thermal power is made using neutron flux instrumentation; however, this instrumentation requires frequent calibration due to changes in the measured flux caused by fuel burn-up, flux pattern changes, and instrumentation drift. To calibrate the nuclear instruments, steam plant calorimetry, a process of performing a heat balance around the nuclear steam supply system, is used. There are four basic elements involved in the calculation of thermal power based on steam plant calorimetry: The mass flow of the feedwater from the power conversion system, the specific enthalpy of that feedwater, the specific enthalpy of the steam delivered to the power conversion system, and other cycle gains and losses. Of these elements, the accuracy of the feedwater mass flow and the feedwater enthalpy, as determined from its temperature and pressure, are typically the largest contributors to the calorimetric calculation uncertainty. Historically, plants have been required to include a margin of 2% in the calculation of the reactor thermal power for the licensed maximum plant output to account for instrumentation uncertainty. The margin is intended to ensure a cushion between operating power and the power for which safety analyses are performed. Use of approved chordal ultrasonic transit-time technology to make the feedwater flow and temperature measurements (in place of traditional differential-pressure- based instruments and resistance temperature detectors [RTDs]) allows for nuclear plant thermal power calculations accurate to 0.3%-0.4% of plant rated power. This improvement in measurement accuracy has allowed many plant operators in the U.S. and around the world to increase plant power output through Measurement Uncertainty Recapture (MUR) up-rates of up to 1.7% of rated power, while also decreasing the probability of significant over-power events. This paper will examine the basic elements involved in calculation of thermal power using ultrasonic transit-time technology and will discuss the criteria for bounding uncertainties associated with each element in order to achieve reactor thermal power calculations to within 0.3% to 0.4%. (authors)

  18. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR)

    SciTech Connect (OSTI)

    Gilbert, B.G.; Richards, R.E.; Reece, W.J.; Gertman, D.I.

    1992-10-01T23:59:59.000Z

    This Reference Guide contains instructions on how to install and use Version 3.5 of the NRC-sponsored Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR). The NUCLARR data management system is contained in compressed files on the floppy diskettes that accompany this Reference Guide. NUCLARR is comprised of hardware component failure data (HCFD) and human error probability (HEP) data, both of which are available via a user-friendly, menu driven retrieval system. The data may be saved to a file in a format compatible with IRRAS 3.0 and commercially available statistical packages, or used to formulate log-plots and reports of data retrieval and aggregation findings.

  19. A New Nuclear Reactor Neutrino Experiment to Measure theta 13

    E-Print Network [OSTI]

    K. Anderson

    2004-02-26T23:59:59.000Z

    An International Working Group has been meeting to discuss ideas for a new Nuclear Reactor Neutrino Experiment at meetings in May 2003 (Alabama), October 2003 (Munich) and plans for March 2004 (Niigata). This White Paper Report on the Motivation and Feasibility of such an experiment is the result of these meetings. After a discussion of the context and opportunity for such an experiment, there are sections on detector design, calibration, overburden and backgrounds, systematic errors, other physics, tunneling issues, safety and outreach. There are 7 appendices describing specific site opportunities.

  20. Iterative methods for solving nonlinear problems of nuclear reactor criticality

    SciTech Connect (OSTI)

    Kuz'min, A. M., E-mail: mephi.kam@mail.ru [National Research Nuclear University MEPhI (Russian Federation)

    2012-12-15T23:59:59.000Z

    The paper presents iterative methods for calculating the neutron flux distribution in nonlinear problems of nuclear reactor criticality. Algorithms for solving equations for variations in the neutron flux are considered. Convergence of the iterative processes is studied for two nonlinear problems in which macroscopic interaction cross sections are functionals of the spatial neutron distribution. In the first problem, the neutron flux distribution depends on the water coolant density, and in the second one, it depends on the fuel temperature. Simple relationships connecting the vapor content and the temperature with the neutron flux are used.

  1. Fast Transient And Spatially Non-Homogenous Accident Analysis Of Two-Dimensional Cylindrical Nuclear Reactor

    SciTech Connect (OSTI)

    Yulianti, Yanti [Dept. of Physics, Universitas Lampung (UNILA), Jl. Sumantri Brojonegor No.1 Bandar Lampung (Indonesia); Dept. of Physics, Institut Teknologi Bandung (ITB), Jl. Ganesha 10 Bandung (Indonesia); Su'ud, Zaki; Waris, Abdul; Khotimah, S. N. [Dept. of Physics, Institut Teknologi Bandung (ITB), Jl. Ganesha 10 Bandung (Indonesia); Shafii, M. Ali [Dept. of Physics, Institut Teknologi Bandung (ITB), Jl. Ganesha 10 Bandung (Indonesia); Dept. of Physics, Universitas Andalas (UNAND), Kampus Limau Manis, Padang, Sumatera Barat (Indonesia)

    2010-12-23T23:59:59.000Z

    The research about fast transient and spatially non-homogenous nuclear reactor accident analysis of two-dimensional nuclear reactor has been done. This research is about prediction of reactor behavior is during accident. In the present study, space-time diffusion equation is solved by using direct methods which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference discretization method is solved by using iterative methods ADI (Alternating Direct Implicit). The indication of accident is decreasing macroscopic absorption cross-section that results large external reactivity. The power reactor has a peak value before reactor has new balance condition. Changing of temperature reactor produce a negative Doppler feedback reactivity. The reactivity will reduce excess positive reactivity. Temperature reactor during accident is still in below fuel melting point which is in secure condition.

  2. 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

  3. ANS Dedicates Reactor as Nuclear Historic Landmark on 50th Anniversary...

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

    Tracy Marc 708.579.8224 ANS Dedicates Reactor as Nuclear Historic Landmark on 50th Anniversary Listen to the audio La Grange Park, IL, - The American Nuclear Society (ANS) honored...

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

    E-Print Network [OSTI]

    Terrani, Kurt Amir

    2010-01-01T23:59:59.000Z

    target subassemblies in a fast reactor core for effectiveof minor actinides in fast reactors [79]. In order to

  5. Gas tagging and cover gas combination for nuclear reactor

    DOE Patents [OSTI]

    Gross, Kenny C. (Lemont, IL); Laug, Matthew T. (Idaho Falls, ID)

    1985-01-01T23:59:59.000Z

    The invention discloses the use of stable isotopes of neon and argon, that are grouped in preselected different ratios one to the other and are then sealed as tags in different cladded nuclear fuel elements to be used in a liquid metal fast breeder reactor. Failure of the cladding of any fuel element allows fission gases generated in the reaction and these tag isotopes to escape and to combine with the cover gas held in the reactor over the fuel elements. The isotopes specifically are Ne.sup.20, Ne.sup.21 and Ne.sup.22 of neon and Ar.sup.36, Ar.sup.38 and Ar.sup.40 of argon, and the cover gas is helium. Serially connected cryogenically operated charcoal beds are used to clean the cover gas and to separate out the tags. The first or cover gas cleanup bed is held between approximately 0.degree. and -25.degree. C. operable to remove the fission gases from the cover gas and tags and the second or tag recovery system bed is held between approximately -170.degree. and -185.degree. C. operable to isolate the tags from the cover gas. Spectrometric analysis further is used to identify the specific tags that are recovered, and thus the specific leaking fuel element. By cataloging the fuel element tags to the location of the fuel elements in the reactor, the location of the leaking fuel element can then be specifically determined.

  6. Monitoring nuclear reactor systems using neural networks and fuzzy logic

    SciTech Connect (OSTI)

    Ikonomopoulos, A.; Tsoukalas, L.H.; Uhrig, R.E. [Tennessee Univ., Knoxville, TN (United States); Mullens, J.A. [Tennessee Univ., Knoxville, TN (United States)]|[Oak Ridge National Lab., TN (United States)

    1991-12-01T23:59:59.000Z

    A new approach is presented that demonstrates the potential of trained artificial neural networks (ANNs) as generators of membership functions for the purpose of monitoring nuclear reactor systems. ANN`s provide a complex-to-simple mapping of reactor parameters in a process analogous to that of measurement. Through such ``virtual measurements`` the value of parameters with operational significance, e.g., control-valve-disk-position, valve-line-up or performance can be determined. In the methodology presented the output of a virtual measuring device is a set of membership functions which independently represent different states of the system. Utilizing a fuzzy logic representation offers the advantage of describing the state of the system in a condensed form, developed through linguistic descriptions and convenient for application in monitoring, diagnostics and generally control algorithms. The developed methodology is applied to the problem of measuring the disk position of the secondary flow control valve of an experimental reactor using data obtained during a start-up. The enhanced noise tolerance of the methodology is clearly demonstrated as well as a method for selecting the actual output. The results suggest that it is possible to construct virtual measuring devices through artificial neural networks mapping dynamic time series to a set of membership functions and thus enhance the capability of monitoring systems. 8 refs., 11 figs., 1 tab.

  7. Selecting a radiation tolerant piezoelectric material for nuclear reactor applications

    SciTech Connect (OSTI)

    Parks, D. A.; Reinhardt, B. T.; Tittmann, B. R. [Department of Engineering Science and Mechanics, Penn State, University Park, PA 16803 (United States)

    2013-01-25T23:59:59.000Z

    Bringing systems for online monitoring of nuclear reactors to fruition has been delayed by the lack of suitable ultrasonic sensors. Recent work has demonstrated the capability of an AlN sensor to perform ultrasonic evaluation in an actual nuclear reactor. Although the AlN demonstrated sustainability, no loss in signal amplitude and d{sub 33} up to a fast and thermal neutron fluence of 1.85 Multiplication-Sign 1018 n/cm{sup 2} and 5.8 Multiplication-Sign 1018 n/cm{sup 2} respectively, no formal process to selecting a suitable sensor material was made. It would be ideal to use first principles approaches to somehow reduce each candidate piezoelectric material to a simple ranking showing directly which materials one should expect to be most radiation tolerant. However, the complexity of the problem makes such a ranking impractical and one must appeal to experimental observations. This should not be of any surprise to one whom is familiar with material science as most material properties are obtained in this manner. Therefore, this work adopts a similar approach, the mechanisms affecting radiation tolerance are discussed and a good engineering sense is used for material qualification of the candidate piezoelectric materials.

  8. Closed Brayton cycle power conversion systems for nuclear reactors :

    SciTech Connect (OSTI)

    Wright, Steven A.; Lipinski, Ronald J.; Vernon, Milton E.; Sanchez, Travis

    2006-04-01T23:59:59.000Z

    This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors, reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at the manufacturers site (Barber-Nichols Inc.) and installed and operated at Sandia. A sufficiently detailed description of the loop is provided in this report along with the design characteristics of the turbo-alternator-compressor set to allow other researchers to compare their results with those measured in the Sandia test-loop. The third task consisted of a validation effort. In this task the test loop was operated and compared with the modeled results to develop a more complete understanding of this electrically heated closed power generation system and to validate the model. The measured and predicted system temperatures and pressures are in good agreement, indicating that the model is a reasonable representation of the test loop. Typical deviations between the model and the hardware results are less than 10%. Additional tests were performed to assess the capability of the Brayton engine to continue to remove decay heat after the reactor/heater is shutdown, to develop safe and effective control strategies, and to access the effectiveness of gas inventory control as an alternative means to provide load following. In one test the heater power was turned off to simulate a rapid reactor shutdown, and the turbomachinery was driven solely by the sensible heat stored in the heater for over 71 minutes without external power input. This is an important safety feature for CBC systems as it means that the closed Brayton loop will keep cooling the reactor without the need for auxiliary power (other than that needed to circulate the waste heat rejection coolant) provided the heat sink is available.

  9. June 28, 2005 France to Be Site of World's First Nuclear Fusion Reactor

    E-Print Network [OSTI]

    June 28, 2005 France to Be Site of World's First Nuclear Fusion Reactor By CRAIG S. SMITH PARIS fusion reactor, an estimated $12 billion project that many scientists see as essential to solving chose the country as the site for the International Thermonuclear Experimental Reactor. Japan, which had

  10. Liquid level, void fraction, and superheated steam sensor for nuclear-reactor cores. [PWR; BWR

    DOE Patents [OSTI]

    Tokarz, R.D.

    1981-10-27T23:59:59.000Z

    This disclosure relates to an apparatus for monitoring the presence of coolant in liquid or mixed liquid and vapor, and superheated gaseous phases at one or more locations within an operating nuclear reactor core, such as pressurized water reactor or a boiling water reactor.

  11. Physics of Nuclear Reactors, March,21 2011 What do we know ?

    E-Print Network [OSTI]

    Danon, Yaron

    Dr. Danon Physics of Nuclear Reactors, March,21 2011 #12;What do we know ? All the information we have is from the media. More reliable; nuclear related information: www.nei.org www.iaea.org THE REST IS INTERPRETATION OF THIS DATA #12;BWR Reactor (Mark I containment) #12;BWR containment in more

  12. A Spouted Bed Reactor Monitoring System for Particulate Nuclear Fuel

    SciTech Connect (OSTI)

    D. S. Wendt; R. L. Bewley; W. E. Windes

    2007-06-01T23:59:59.000Z

    Conversion and coating of particle nuclear fuel is performed in spouted (fluidized) bed reactors. The reactor must be capable of operating at temperatures up to 2000°C in inert, flammable, and coating gas environments. The spouted bed reactor geometry is defined by a graphite retort with a 2.5 inch inside diameter, conical section with a 60° included angle, and a 4 mm gas inlet orifice diameter through which particles are removed from the reactor at the completion of each run. The particles may range from 200 µm to 2 mm in diameter. Maintaining optimal gas flow rates slightly above the minimum spouting velocity throughout the duration of each run is complicated by the variation of particle size and density as conversion and/or coating reactions proceed in addition to gas composition and temperature variations. In order to achieve uniform particle coating, prevent agglomeration of the particle bed, and monitor the reaction progress, a spouted bed monitoring system was developed. The monitoring system includes a high-sensitivity, low-response time differential pressure transducer paired with a signal processing, data acquisition, and process control unit which allows for real-time monitoring and control of the spouted bed reactor. The pressure transducer is mounted upstream of the spouted bed reactor gas inlet. The gas flow into the reactor induces motion of the particles in the bed and prevents the particles from draining from the reactor due to gravitational forces. Pressure fluctuations in the gas inlet stream are generated as the particles in the bed interact with the entering gas stream. The pressure fluctuations are produced by bulk movement of the bed, generation and movement of gas bubbles through the bed, and the individual motion of particles and particle subsets in the bed. The pressure fluctuations propagate upstream to the pressure transducer where they can be monitored. Pressure fluctuation, mean differential pressure, gas flow rate, reactor operating temperature data from the spouted bed monitoring system are used to determine the bed operating regime and monitor the particle characteristics. Tests have been conducted to determine the sensitivity of the monitoring system to the different operating regimes of the spouted particle bed. The pressure transducer signal response was monitored over a range of particle sizes and gas flow rates while holding bed height constant. During initial testing, the bed monitoring system successfully identified the spouting regime as well as when particles became interlocked and spouting ceased. The particle characterization capabilities of the bed monitoring system are currently being tested and refined. A feedback control module for the bed monitoring system is currently under development. The feedback control module will correlate changes in the bed response to changes in the particle characteristics and bed spouting regime resulting from the coating and/or conversion process. The feedback control module will then adjust the gas composition, gas flow rate, and run duration accordingly to maintain the bed in the desired spouting regime and produce optimally coated/converted particles.

  13. Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

    DOE Patents [OSTI]

    Youchison, Dennis L. (Albuquerque, NM); Williams, Brian E. (Pacoima, CA); Benander, Robert E. (Pacoima, CA)

    2011-03-01T23:59:59.000Z

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  14. Methods for manufacturing porous nuclear fuel elements for high-temperature gas-cooled nuclear reactors

    DOE Patents [OSTI]

    Youchison, Dennis L. (Albuquerque, NM); Williams, Brian E. (Pocoima, CA); Benander, Robert E. (Pacoima, CA)

    2010-02-23T23:59:59.000Z

    Methods for manufacturing porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's). Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, a thin coating of nuclear fuel may be deposited inside of a highly porous skeletal structure made, for example, of reticulated vitreous carbon foam.

  15. Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

    DOE Patents [OSTI]

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2013-09-03T23:59:59.000Z

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  16. Vital area identification for U.S. Nuclear Regulatory Commission nuclear power reactor licensees and new reactor applicants.

    SciTech Connect (OSTI)

    Whitehead, Donnie Wayne; Varnado, G. Bruce

    2008-09-01T23:59:59.000Z

    U.S. Nuclear Regulatory Commission nuclear power plant licensees and new reactor applicants are required to provide protection of their plants against radiological sabotage, including the placement of vital equipment in vital areas. This document describes a systematic process for the identification of the minimum set of areas that must be designated as vital areas in order to ensure that all radiological sabotage scenarios are prevented. Vital area identification involves the use of logic models to systematically identify all of the malicious acts or combinations of malicious acts that could lead to radiological sabotage. The models available in the plant probabilistic risk assessment and other safety analyses provide a great deal of the information and basic model structure needed for the sabotage logic model. Once the sabotage logic model is developed, the events (or malicious acts) in the model are replaced with the areas in which the events can be accomplished. This sabotage area logic model is then analyzed to identify the target sets (combinations of areas the adversary must visit to cause radiological sabotage) and the candidate vital area sets (combinations of areas that must be protected against adversary access to prevent radiological sabotage). Any one of the candidate vital area sets can be selected for protection. Appropriate selection criteria will allow the licensee or new reactor applicant to minimize the impacts of vital area protection measures on plant safety, cost, operations, or other factors of concern.

  17. Measuring of fissile isotopes partial antineutrino spectra in direct experiment at nuclear reactor

    E-Print Network [OSTI]

    V. V. Sinev

    2009-02-22T23:59:59.000Z

    The direct measuring method is considered to get nuclear reactor antineutrino spectrum. We suppose to isolate partial spectra of the fissile isotopes by using the method of antineutrino spectrum extraction from the inverse beta decay positron spectrum applied at Rovno experiment. This admits to increase the accuracy of partial antineutrino spectra forming the total nuclear reactor spectrum. It is important for the analysis of the reactor core fuel composition and could be applied for non-proliferation purposes.

  18. The harmony between nuclear reactions and nuclear reactor structures and systems

    SciTech Connect (OSTI)

    Popa-Simil, L. [LAVM LLC, Los Alamos, NM (United States)

    2012-07-01T23:59:59.000Z

    Advanced nuclear energy is one extremely viable approach for achieving the required goals. With its extraordinarily high energy density (both, per unit mass and per unit volume), it produces over seven orders of magnitude less waste than fossil fuels per unit of energy generated. Applying nano-technologies to nuclear reactors could potentially produce the extraordinary performance required. The actual nuclear reactors lack of performances, the complexity and hazard of the fuel cycle are in part due to the lack of understanding of the nature's laws related to energy distribution applied to fission products, and in part to the current technologic capabilities that make the economical optimum. In order to produce the desired increase of performances a novel multi-scale multi-physics and engineering approach have been developed, starting from the nuclear reactions involved, analyzing in detail the key features and requirements of the 'key players' in the process (neutrons, compound nucleus, fission products, transmutation products, decay radiation), the consequences of their interaction with matter. That complex interaction generates new reactions and new key-players (knock-on electrons, photons, phonons) that further interact with the matter represented by the nuclear fuel, cladding, cooling agents, structural materials and control systems. The understanding of this complexity of problems from fm-ps scale up to macro-system and mitigating all the requirements drives to that desired harmony that provides a safe energy delivery. (authors)

  19. Closing nuclear fuel cycle with fast reactors: problems and prospects

    SciTech Connect (OSTI)

    Shadrin, A.; Dvoeglazov, K.; Ivanov, V. [Bochvar Institute - VNIINM, Moscow (Russian Federation)

    2013-07-01T23:59:59.000Z

    The closed nuclear fuel cycle (CNFC) with fast reactors (FR) is the most promising way of nuclear energetics development because it prevents spent nuclear fuel (SNF) accumulation and minimizes radwaste volume due to minor actinides (MA) transmutation. CNFC with FR requires the elaboration of safety, environmentally acceptable and economically effective methods of treatment of SNF with high burn-up and low cooling time. The up-to-date industrially implemented SNF reprocessing technologies based on hydrometallurgical methods are not suitable for the reprocessing of SNF with high burn-up and low cooling time. The alternative dry methods (such as electrorefining in molten salts or fluoride technologies) applicable for such SNF reprocessing have not found implementation at industrial scale. So the cost of SNF reprocessing by means of dry technologies can hardly be estimated. Another problem of dry technologies is the recovery of fissionable materials pure enough for dense fuel fabrication. A combination of technical solutions performed with hydrometallurgical and dry technologies (pyro-technology) is proposed and it appears to be a promising way for the elaboration of economically, ecologically and socially accepted technology of FR SNF management. This paper deals with discussion of main principle of dry and aqueous operations combination that probably would provide safety and economic efficiency of the FR SNF reprocessing. (authors)

  20. Method of nuclear reactor control using a variable temperature load dependent set point

    SciTech Connect (OSTI)

    Kelly, J.J.; Rambo, G.E.

    1982-04-27T23:59:59.000Z

    A method and apparatus for controlling a nuclear reactor in response to a variable average reactor coolant temperature set point is disclosed. The set point is dependent upon percent of full power load demand. A manually-actuated ''droop mode'' of control is provided whereby the reactor coolant temperature is allowed to drop below the set point temperature a predetermined amount wherein the control is switched from reactor control rods exclusively to feedwater flow.

  1. NGNP Project Regulatory Gap Analysis for Modular HTGRs

    SciTech Connect (OSTI)

    Wayne Moe

    2011-09-01T23:59:59.000Z

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

  2. International Topical Meeting on Nuclear Reactor Thermalhydraulics, NURETH-15 NURETH15-xxx Pisa, Italy, May 12-15, 2013

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    The 15th International Topical Meeting on Nuclear Reactor Thermalhydraulics, NURETH-15 NURETH15-xxx technologies in the context of generation IV nuclear power reactors. In order to improve electric efficiency during last years as a possible energy conversion cycle for Sodium nuclear Fast Reactors (SFRs) [1

  3. Record of Cycling Operation of the Natural Nuclear Reactor in the Oklo/Okelobondo Area in Gabon

    E-Print Network [OSTI]

    Record of Cycling Operation of the Natural Nuclear Reactor in the Oklo/Okelobondo Area in Gabon A billion yr old Oklo natural nuclear reactor. In addition to elevated abundances of fission-produced Zr, Ce nuclear chain reaction was predicted by Kuroda [1] 20 years before the remnants of the natural reactor

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

    E-Print Network [OSTI]

    Terrani, Kurt Amir

    2010-01-01T23:59:59.000Z

    capacity and operating efficiency of nuclear plants [31,operating efficiency of nuclear plants in the past decades.cost of the fuel Nuclear Plant Capacity Factor Nuclear

  5. Nuclear reactor fuel element having improved heat transfer

    DOE Patents [OSTI]

    Garnier, J.E.; Begej, S.; Williford, R.E.; Christensen, J.A.

    1982-03-03T23:59:59.000Z

    A nuclear reactor fuel element having improved heat transfer between fuel material and cladding is described. The element consists of an outer cladding tube divided into an upper fuel section containing a central core of fissionable or mixed fissionable and fertile fuel material, slightly smaller in diameter than the inner surface of the cladding tube and a small lower accumulator section, the cladding tube being which is filled with a low molecular weight gas to transfer heat from fuel material to cladding during irradiation. A plurality of essentially vertical grooves in the fuel section extend downward and communicate with the accumulator section. The radial depth of the grooves is sufficient to provide a thermal gradient between the hot fuel surface and the relatively cooler cladding surface to allow thermal segregation to take place between the low molecular weight heat transfer gas and high molecular weight fission product gases produced by the fuel material during irradiation.

  6. Nuclear reactor control rod with uniformly changeable axial worth

    SciTech Connect (OSTI)

    Freeman, T.R.

    1989-04-11T23:59:59.000Z

    A control rod is described for use in a nuclear reactor core to provide xenon compensation, comprising: (a) an elongated inner cylindrical member having a lower end; and (b) an elongated outer cylindrical member surrounding the inner member and having a lower end with concentrically-arranged inner and outer edge portions defined thereon; (c) each of the members being composed of alternating poison and nonpoison regions; (d) the inner member being axially movable relative to the outer member to adjust the degree to which the poison regions of the members overlap with the nonpoison regions thereof and thereby change the overall worth of the rod; and (e) the lower end of the inner member having defined thereon a radially outwardly projecting ledge for supporting in a rest relationship thereon the lower end of the outer member at only its inner edge portion for retaining the outer member about the inner member.

  7. Reactor and shielding design implications of clustering nuclear thermal rockets

    SciTech Connect (OSTI)

    Buksa, J.J.; Houts, M.G. (Los Alamos National Laboratory, NM (United States))

    1992-07-01T23:59:59.000Z

    This paper examines design considerations in the context of engine-out accidents in clustered nuclear-thermal rocket stages, and an accident-management protocol is devised. Safety and performance issues are considered in the light of designs for the reactor and shielding elements of ROVER/NERVA-type engines. The engine-out management process involves: phase one, in which sufficient propulsive power is guaranteed for mission completion; and phase two, in which engine failure is isolated and not allowed to propagate to other engines or to the spacecraft. Phase-one designs can employ spare engines, throttled engines, and/or long-burning engines. Phase-two safety concepts can include techniques for cooling or jettisoning the failed engines. Engine-out management philosophies are shown to be shaped by a combination of safety and mission-trajectory requirements. 6 refs.

  8. Improved gas tagging and cover gas combination for nuclear reactor

    DOE Patents [OSTI]

    Gross, K.C.; Laug, M.T.

    1983-09-26T23:59:59.000Z

    The invention discloses the use of stable isotopes of neon and argon, sealed as tags in different cladding nuclear fuel elements to be used in a liquid metal fast breeder reactor. Cladding failure allows fission gases and these tag isotopes to escape and to combine with the cover gas. The isotopes are Ne/sup 20/, Ne/sup 21/ and Ne/sup 22/ and Ar/sup 36/, Ar/sup 38/ and Ar/sup 40/, and the cover gas is He. Serially connected cryogenically operated charcoal beds are used to clean the cover gas and to separate out the tags. The first or cover gas cleanup bed is held between 0 and -25/sup 0/C to remove the fission gases from the cover gas and tags, and the second or tag recovery system bed between -170 and -185/sup 0/C to isolate the tags from the cover gas. Spectrometric analysis is used to identify the specific tags that are recovered, and thus the specific leaking fuel element. By cataloging the fuel element tags to the location of the fuel elements in the reactor, the location of the leaking fuel element can then be determined.

  9. Modeling of thermophoretic deposition of aerosols in nuclear reactor containments

    SciTech Connect (OSTI)

    Fernandes, A.; Loyalka, S.K. [Univ. of Missouri, Columbia, MO (United States)

    1996-12-01T23:59:59.000Z

    Aerosol released in postulated or real nuclear reactor accidents can deposit on containment surfaces via motion induced by temperature gradients in addition to the motion due to diffusion and gravity. The deposition due to temperature gradients is known as thermophoretic deposition, and it is currently modeled in codes such as CONTAIN in direct analogy with heat transfer, but there have been questions about such analogies. This paper focuses on a numerical solution of the particle continuity equation in laminar flow condition characteristics of natural convection. First, the thermophoretic deposition rate is calculated as a function of the Prandtl and Schmidt numbers, the thermophoretic coefficient K, and the temperature difference between the atmosphere and the wall. Then, the cases of diffusion alone and a boundary-layer approximation (due to Batchelor and Shen) to the full continuity equation are considered. It is noted that an analogy with heat transfer does not hold, but for the circumstances considered in this paper, the deposition rates from the diffusion solution and the boundary-layer approximation can be added to provide reasonably good agreement (maximum deviation 30%) with the full solution of the particle continuity equation. Finally, correlations useful for implementation in the reactor source term codes are provided.

  10. Nuclear reactor cooling system decontamination reagent regeneration. [PWR; BWR

    DOE Patents [OSTI]

    Anstine, L.D.; James, D.B.; Melaika, E.A.; Peterson, J.P. Jr.

    1980-06-06T23:59:59.000Z

    An improved method for decontaminating the coolant system of water-cooled nuclear power reactors and for regenerating the decontamination solution is described. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution.

  11. Self-actuating and locking control for nuclear reactor

    DOE Patents [OSTI]

    Chung, Dong K. (Chatsworth, CA)

    1982-01-01T23:59:59.000Z

    A self-actuating, self-locking flow cutoff valve particularly suited for use in a nuclear reactor of the type which utilizes a plurality of fluid support neutron absorber elements to provide for the safe shutdown of the reactor. The valve comprises a substantially vertical elongated housing and an aperture plate located in the housing for the flow of fluid therethrough, a substantially vertical elongated nozzle member located in the housing and affixed to the housing with an opening in the bottom for receiving fluid and apertures adjacent a top end for discharging fluid. The nozzle further includes two sealing means, one located above and the other below the apertures. Also located in the housing and having walls surrounding the nozzle is a flow cutoff sleeve having a fluid opening adjacent an upper end of the sleeve, the sleeve being moveable between an upper open position wherein the nozzle apertures are substantially unobstructed and a closed position wherein the sleeve and nozzle sealing surfaces are mated such that the flow of fluid through the apertures is obstructed. It is a particular feature of the present invention that the valve further includes a means for utilizing any increase in fluid pressure to maintain the cutoff sleeve in a closed position. It is another feature of the invention that there is provided a means for automatically closing the valve whenever the flow of fluid drops below a predetermined level.

  12. Numerical analysis of turbulent heat transfer in a nuclear reactor coolant channel

    E-Print Network [OSTI]

    Garrard, Clarence William

    1965-01-01T23:59:59.000Z

    NUMERICAL ANALYSIS OF TURBULENT HEAT TRANSFER IN A NUCLEAR REACTOR COOLANT CHANNEL A Thesis Clarence William Garrard, Jr. Submitted to the Graduate College of the Texas A&M University in partial fulfillment of' the requirements for the degree... of' MASTER OF SC1ENCE May, 1965 Ma)or Subject Nuclear Engineering NUMERICAL ANALYSIS OF TURBULENT HEAT TRANSFER 1N A NUCLEAR REACTOR COOLANT CHANNEL A Thesis By Clarence William Garrard, Jr. Approved as to style and content by; Head...

  13. SciTech Connect: Nuclear power reactor instrumentation systems...

    Office of Scientific and Technical Information (OSTI)

    of Publication: United States Language: English Subject: N79400* --Reactors--Reactor Control Systems; N46110 -- Instrumentation--Radiation Detection Instruments--General...

  14. Underground collocation of nuclear power plant reactors and repository to facilitate the post-renaissance expansion of nuclear power

    SciTech Connect (OSTI)

    Myers, Carl W [Los Alamos National Laboratory; Elkins, Ned Z [Los Alamos National Laboratory

    2008-01-01T23:59:59.000Z

    Underground collocation of nuclear power reactors and the nuclear waste management facilities supporting those reactors, termed an underground nuclear park (UNP), appears to have several advantages compared to the conventional approach to siting reactors and waste management facilities. These advantages include the potential to lower reactor capital and operating cost, lower nuclear waste management cost, and increase margins of physical security and safety. Envirorunental impacts related to worker health, facility accidents, waste transportation, and sabotage and terrorism appear to be lower for UNPs compared to the current approach. In-place decommissioning ofUNP reactors appears to have cost, safety, envirorunental and waste disposal advantages. The UNP approach has the potential to lead to greater public acceptance for the deployment of new power reactors. Use of the UNP during the post-nuclear renaissance time frame has the potential to enable a greater expansion of U.S. nuclear power generation than might otherwise result. Technical and economic aspects of the UNP concept need more study to determine the viability of the concept.

  15. A Generalized Adjoint Framework for Sensitivity and Global Error Estimation in Time-Dependent Nuclear Reactor Simulations 1

    E-Print Network [OSTI]

    Anitescu, Mihai

    -Dependent Nuclear Reactor Simulations 1 H. F. Striplinga, , M. Anitescub , M. L. Adamsa aNuclear Engineering Bateman and transport equations, which govern the time-dependent neutronic behavior of a nuclear reactor framework for computing the adjoint variable to nuclear engineering problems gov- erned by a set

  16. An underground nuclear power station using self-regulating heat-pipe controlled reactors

    DOE Patents [OSTI]

    Hampel, V.E.

    1988-05-17T23:59:59.000Z

    A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast- acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor. 5 figs.

  17. Underground nuclear power station using self-regulating heat-pipe controlled reactors

    DOE Patents [OSTI]

    Hampel, Viktor E. (Pleasanton, CA)

    1989-01-01T23:59:59.000Z

    A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working flud in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor.

  18. First Step to Spur U.S. Manufacturing of Small Modular Nuclear Reactors |

    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 onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directed off Energy.gov. AreSecond SolarExternal Technical6-OPAM2FederalEnergyProgramFirstDepartment of

  19. Advancing Small Modular Reactors: How We're Supporting Next-Gen Nuclear

    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 onYouTube YouTube Note: Since the YouTube platform is alwaysISO 50001EnergyNewsletter AdvancedWindow andCivilEnergy

  20. Nuclear reactor control rod having a reduced worth tip

    SciTech Connect (OSTI)

    Doshi, P.K.; Wilson, J.F.

    1986-11-25T23:59:59.000Z

    This patent describes a nuclear reactor having a fuel assembly and a control rod moveable in and out for controlling the reactivity of the reactor, the control rod comprising: (a) an elongated tubular cladding; (b) means for closing the opposite ends of the cladding; (c) first pellets of a first type disposed within the cladding in an end-to-end relationship; and (d) second pellets of a second type disposed within the cladding in an end-to-end relationship; (e) one of the first and second types of pellets being formed of a material having a generally high neutron absorbing capacity; (f) the other of the first and second types of pellets being formed of an inert material having a generally low neutron absorbing capacity; (g) the pellets of the first and second types thereof being cylindrical with their respective diameters being generally equal; (h) the inert pellets being interspaced between the high neutron absorbing pellets at a lower end portion of the cladding with the remaining portion of the cladding above the lower end portion containing only the high neutron absorbing pellets; and (i) the axial heights of one of the first and second pluralities of pellets of the first and second types located at the lower end portion of the cladding progressively varying from pellet to pellet, the axial heights of the other of the first and second pellets of the first and second types located at the lower end portion of the cladding are generally equal from pellet to pellet, so as to produce an improved reduced worth tip at the lower end portion of the cladding.

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

    E-Print Network [OSTI]

    Terrani, Kurt Amir

    2010-01-01T23:59:59.000Z

    electricity generation capacity and operating efficiency of nuclear plants [Nuclear Plant Capacity Factor Nuclear Electricity Generationelectricity generation capacity and operating efficiency of nu- clear plants [

  2. Role of fast reactor and its cycle to reduce nuclear waste burden

    SciTech Connect (OSTI)

    Arie, Kazuo; Oomori, Takashi; Okita, Takeshi [Toshiba Corporation, 8, Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan); Kawashima, Masatoshi [Toshiba Nuclear Engineering Services Corporation, 8, Shinsugita-cho, Isogo-ku, Yokohama, 235-8523 (Japan); Kotake, Shoji [The Japan Atomic Power Company, 1-1, Kanda-Mitoshiro-cho, Chiyoda-ku, Tokyo 101-0053 (Japan); Fuji-ie, Yoichi [Nuclear Salon Fuji-ie, 1-11-10, Yushima, Bunkyo-ku, Tokyo 113-0034 (Japan)

    2013-07-01T23:59:59.000Z

    The role of the metal fuel fast reactor with recycling of actinides and the five long-lived fission products based on the concept of the Self-Consistent Nuclear Energy System has been examined by evaluating the reduction of nuclear wastes during the transition period to this reactor system. The evaluation was done in comparison to an LWR once-through case and a conventional actinide recycling oxide fast reactor. As a result, it is quantitatively clarified that a metal fuel fast reactor with actinide and the five long-lived fission products (I{sup 129}, Tc{sup 99}, Zr{sup 93}, Cs{sup 135} and Sn{sup 126}) recycling could play a significant role in reducing the nuclear waste burden including the current LWR wastes. This can be achieved by using a fast neutron spectrum reactor enhanced with metal fuel that brings high capability as a 'waste burner'. (authors)

  3. 1 hour, 59 minutes ago President Jacques Chirac announced plans to build a prototype fourth-generation nuclear reactor by 2020 as well as symbolic targets

    E-Print Network [OSTI]

    -generation nuclear reactor by 2020 as well as symbolic targets for cutting France's reliance on oil in the coming and is conducting research into several new models of nuclear reactor. Business leaders in the French energy sector-generation nuclear reactor 1/5/06 3:19 PMPrint Story: France to develop fourth-generation nuclear reactor on Yahoo

  4. Sodium Cooled Fast Reactors and the Pyro-Process: Conversion of Nuclear Waste into a Fuel Source

    E-Print Network [OSTI]

    Belanger, David P.

    1 Sodium Cooled Fast Reactors and the Pyro-Process: Conversion of Nuclear Waste into a Fuel Source renewed interest amongst the nuclear science community as the debate over nuclear waste has increased .................................................................................27 2.1.2 Waste Minimization

  5. Independent Confirmatory Survey Report for the University of Arizona Nuclear Reactor Laboratory, Tucson, Arizona

    SciTech Connect (OSTI)

    Nick A. Altic

    2011-11-11T23:59:59.000Z

    The University of Arizona (University) research reactor is a TRIGA swimming pool type reactor designed by General Atomics and constructed at the University in 1958. The reactor first went into operation in December of 1958 under U.S. Nuclear Regulatory Commission (NRC) license R-52 until final shut down on May 18, 2010. Initial site characterization activities were conducted in February 2009 during ongoing reactor operations to assess the radiological status of the Nuclear Reactor Laboratory (NRL) excluding the reactor tank, associated components, and operating systems. Additional post-shutdown characterization activities were performed to complete characterization activities as well as verify assumptions made in the Decommissioning Plan (DP) that were based on a separate activation analysis (ESI 2009 and WMG 2009). Final status survey (FSS) activities began shortly after the issuance of the FSS plan in May 2011. The contractor completed measurement and sampling activities during the week of August 29, 2011.

  6. Power conversion system design for supercritical carbon dioxide cooled indirect cycle nuclear reactors

    E-Print Network [OSTI]

    Gibbs, Jonathan Paul

    2008-01-01T23:59:59.000Z

    The supercritical carbon dioxide (S-CO?) cycle is a promising advanced power conversion cycle which couples nicely to many Generation IV nuclear reactors. This work investigates the power conversion system design and ...

  7. Examination of offsite radiological emergency protective measures for nuclear reactor accidents involving core melt

    E-Print Network [OSTI]

    Aldrich, David C.

    1979-01-01T23:59:59.000Z

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted ...

  8. The development of a remote monitoring system for the Nuclear Science Center reactor

    E-Print Network [OSTI]

    Jiltchenkov, Dmitri Victorovich

    2002-01-01T23:59:59.000Z

    With funding provided by Nuclear Energy Research Initiative (NERI), design of Secure, Transportable, Autonomous Reactors (STAR) to aid countries with insufficient energy supplies is underway. The development of a new monitoring system that allows...

  9. Space nuclear-power reactor design based on combined neutronic and thermal-fluid analyses

    SciTech Connect (OSTI)

    Koenig, D.R.; Gido, R.G.; Brandon, D.I.

    1985-01-01T23:59:59.000Z

    The design and performance analysis of a space nuclear-power system requires sophisticated analytical capabilities such as those developed during the nuclear rocket propulsion (Rover) program. In particular, optimizing the size of a space nuclear reactor for a given power level requires satisfying the conflicting requirements of nuclear criticality and heat removal. The optimization involves the determination of the coolant void (volume) fraction for which the reactor diameter is a minimum and temperature and structural limits are satisfied. A minimum exists because the critical diameter increases with increasing void fraction, whereas the reactor diameter needed to remove a specified power decreases with void fraction. The purpose of this presentation is to describe and demonstrate our analytical capability for the determination of minimum reactor size. The analysis is based on combining neutronic criticality calculations with OPTION-code thermal-fluid calculations.

  10. Print this article Close This Window EU OKs India joining ITER nuclear reactor project

    E-Print Network [OSTI]

    Print this article Close This Window EU OKs India joining ITER nuclear reactor project Fri Dec 2 trademarks and trademarks of the Reuters group of companies around the world. Close This Window 12/2/05 4

  11. Passive cooling system for liquid metal cooled nuclear reactors with backup coolant flow path

    DOE Patents [OSTI]

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

    1993-01-01T23:59:59.000Z

    A liquid metal cooled nuclear fission reactor plant having a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during reactor shutdown, or heat produced during a mishap. This reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary system when rendered inoperable.

  12. A Computer Program Predicting Steady-State Performance of a Nuclear Research Reactor's Cooling System

    SciTech Connect (OSTI)

    Kamel Sidi Ali [Nuclear Research Center of Birine (Algeria)

    2002-07-01T23:59:59.000Z

    The performances of a nuclear reactor are directly affected by its cooling system, especially when it uses wet towers to evacuate the heat generated in the nuclear reactor core. Failure of the cooling system can yield very serious damages to most of the components of the nuclear reactor core. In this work, a computer program simulating the thermal behavior of a nuclear research reactor's cooling system is presented. Starting from the proposed start-up data of the reactor, the program predicts the cooling capacity of the nuclear reactor while taking into account the current climate conditions and also monitors the behavior of the thermal equipment involved in this process and this for different levels of power. The proposed simulation is based on a set of heat transfer equations representing all the equipment making up the cooling system up to the nuclear reactor core. Owing to the proposed inter-connected set of equations used to predict the thermal behaviour of the system, this program allows the user to modify at will a specified parameter and study the induced resulting effects on the rest of the system. The computer program developed has been experimentally validated on an operational system generating 6.8 MW and the obtained results are in good agreement with experiment. The results produced by the program concern the capacity of the cooling system to evacuate all the heat generated in the nuclear reactor core while taking into account the current climate conditions, the determination of the optimal number of thermal equipment that need to be engaged, the monitoring of the reactor core's entry end exit temperatures as well as the temperatures of all the components of the cooling system. Moreover, the program gives all the characteristics of air at the exit of the cooling towers and the loss of water due to the cooling process. (authors)

  13. Modelling the Electron Beam Welding of Nuclear Reactor Pressure Vessel Steel

    E-Print Network [OSTI]

    Cambridge, University of

    Modelling the Electron Beam Welding of Nuclear Reactor Pressure Vessel Steel Christopher J. Duffy fabrication of thick-section steel for critical components such as reactor pressure vessels. Electron beam weld tests performed by Rolls-Royce and The Welding Institute of SA 508 Grade 3 and SA 508 Grade 4N

  14. Capillary-Pumped Passive Reactor Concept for Space Nuclear Power

    SciTech Connect (OSTI)

    Dr. Thomas F. Lin; Dr. Thomas G. Hughes; Christopher G. Miller

    2008-05-30T23:59:59.000Z

    To develop the passively-cooled space reactor concept using the capillary-induced lithium flow, since molten lithium possesses a very favorable surface tension characteristic. In space where the gravitational field is minimal, the gravity-assisted natural convection cooling is not effective nor an option for reactor heat removal, the capillary induced cooling becomes an attractive means of providing reactor cooling.

  15. The design and construction of a 10-amplifier analog computer with provisions for nuclear reactor simulation

    E-Print Network [OSTI]

    Cox, James Robert

    1959-01-01T23:59:59.000Z

    THE DESIGN AND CONSTRUCTION OF A 10-AMPLIFIER ANALOG COMPUTER WITH PROVISIONS FOR NUCLEAR REACTOR SIMULATION A Thesis by James Robert Cox Submitted to the Graduate School of the Agricultural and Mechanical College of Texas in partial... fulfillment of the requirements for the degree of MASTER OF SCIENCE August 1959 Major Subject: Ele ctr ical Engines r ing THE DESIGN AND CONSTRUCTION OF A 10-AMPLIFIER ANALOG COMPUTER WITH PROVISIONS FOR NUCLEAR REACTOR SIMULATION A The s is by Jame...

  16. Considerations for absorber analysis by reactivity measurements in a nuclear reactor

    E-Print Network [OSTI]

    Kaufman, Jay Leslie

    1959-01-01T23:59:59.000Z

    CONSIUERATIONS FOR ASSOESER ANALYSIS bY REACTIVITY NEASURENENTS IN A NUCLEAR REACTOR JAY LESLXE EAUFNAN Submitted to the Graduate School of tha Agricultural and Mechanical College of Texas in partial fulfillment of the requirements... for the degree of NASTER OF SCIENCE August 1959 Ma)or Sub)ect t Fhysics CONSIDERATIONS POR ABSORBER ANALYSIS BY REACTIVITY MEASUREHENTS IN A NUCLEAR REACTOR A Thssia By JA. Y LESLLE EAUPNAN Approved as to style and content by: (Chainsan of Coenittss...

  17. Hybrid nuclear reactor grey rod to obtain required reactivity worth

    DOE Patents [OSTI]

    Miller, John V. (Munhall, PA); Carlson, William R. (Scott Township, Allegheny County, PA); Yarbrough, Michael B. (Hempfield Township, Westmoreland County, PA)

    1991-01-01T23:59:59.000Z

    Hybrid nuclear reactor grey rods are described, wherein geometric combinations of relatively weak neutron absorber materials such as stainless steel, zirconium or INCONEL, and relatively strong neutron absorber materials, such as hafnium, silver-indium cadmium and boron carbide, are used to obtain the reactivity worths required to reach zero boron change load follow. One embodiment includes a grey rod which has combinations of weak and strong neutron absorber pellets in a stainless steel cladding. The respective pellets can be of differing heights. A second embodiment includes a grey rod with a relatively thick stainless steel cladding receiving relatively strong neutron absorber pellets only. A third embodiment includes annular relatively weak netron absorber pellets with a smaller diameter pellet of relatively strong absorber material contained within the aperture of each relatively weak absorber pellet. The fourth embodiment includes pellets made of a homogeneous alloy of hafnium and a relatively weak absorber material, with the percentage of hafnium chosen to obtain the desired reactivity worth.

  18. Eastern Europe Research Reactor Initiative nuclear education and training courses - Current activities and future challenges

    SciTech Connect (OSTI)

    Snoj, L. [Josef Stefan Inst., Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Sklenka, L.; Rataj, J. [Dept. of Nuclear Reactor, Czech Technical Univ. in Prague, V Holesovickach 2, 180 00 Prague 8 (Czech Republic); Boeck, H. [Vienna Univ. of Technology/Atominstitut, Stadionallee 2, 1020 Vienna (Austria)

    2012-07-01T23:59:59.000Z

    The Eastern Europe Research Reactor Initiative was established in January 2008 to enhance cooperation between the Research Reactors in Eastern Europe. It covers three areas of research reactor utilisation: irradiation of materials and fuel, radioisotope production, neutron beam experiments, education and training. In the field of education and training an EERRI training course was developed. The training programme has been elaborated with the purpose to assist IAEA Member States, which consider building a research reactor (RR) as a first step to develop nuclear competence and infrastructure in the Country. The major strength of the reactor is utilisation of three different research reactors and a lot of practical exercises. Due to high level of adaptability, the course can be tailored to specific needs of institutions with limited or no access to research reactors. (authors)

  19. CONSTRUCTION OF WEB-ACCESSIBLE MATERIALS HANDBOOK FORGENERATION IV NUCLEAR REACTORS

    SciTech Connect (OSTI)

    Ren, Weiju [ORNL

    2005-01-01T23:59:59.000Z

    The development of a web-accessible materials handbook in support of the materials selection and structural design for the Generation IV nuclear reactors is being planned. Background of the reactor program is briefly introduced. Evolution of materials handbooks for nuclear reactors over years is reviewed in light of the trends brought forth by the rapid advancement in information technologies. The framework, major features, contents, and construction considerations of the web-accessible Gen IV Materials Handbook are discussed. Potential further developments and applications of the handbook are also elucidated.

  20. Application of an exact model matching technique to coupled-core nuclear reactor control

    SciTech Connect (OSTI)

    Tzafestas, S.G.; Chrysochoides, N.G.; Rokkos, K.

    1984-08-01T23:59:59.000Z

    In this Note the control problem of linearized coupled-core multivariable nuclear reactors is treated by using a recent exact model matching technique in the frequency domain. The case of state feedback control is first considered and then the results are used where only the output variables of the reactor are available for feedback. A numerical example of a three coupled-core nuclear reactor model with one delayed neutron group for each core and short neutron travel time between cores is included.

  1. SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

    ScienceCinema (OSTI)

    None

    2014-03-11T23:59:59.000Z

    U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in a remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.

  2. SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

    SciTech Connect (OSTI)

    None

    2013-09-25T23:59:59.000Z

    U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in a remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.

  3. Symmetric modular torsatron

    DOE Patents [OSTI]

    Rome, J.A.; Harris, J.H.

    1984-01-01T23:59:59.000Z

    A fusion reactor device is provided in which the magnetic fields for plasma confinement in a toroidal configuration is produced by a plurality of symmetrical modular coils arranged to form a symmetric modular torsatron referred to as a symmotron. Each of the identical modular coils is helically deformed and comprise one field period of the torsatron. Helical segments of each coil are connected by means of toroidally directed windbacks which may also provide part of the vertical field required for positioning the plasma. The stray fields of the windback segments may be compensated by toroidal coils. A variety of magnetic confinement flux surface configurations may be produced by proper modulation of the winding pitch of the helical segments of the coils, as in a conventional torsatron, winding the helix on a noncircular cross section and varying the poloidal and radial location of the windbacks and the compensating toroidal ring coils.

  4. Nuclear reactor having a polyhedral primary shield and removable vessel insulation

    DOE Patents [OSTI]

    Ekeroth, D.E.; Orr, R.

    1993-12-07T23:59:59.000Z

    A nuclear reactor is provided having a generally cylindrical reactor vessel disposed within an opening in a primary shield. The opening in the primary shield is defined by a plurality of generally planar side walls forming a generally polyhedral-shaped opening. The reactor vessel is supported within the opening in the primary shield by reactor vessel supports which are in communication and aligned with central portions of some of the side walls. The reactor vessel is connected to the central portions of the reactor vessel supports. A thermal insulation polyhedron formed from a plurality of slidably insertable and removable generally planar insulation panels substantially surrounds at least a portion of the reactor vessel and is disposed between the reactor vessel and the side walls of the primary shield. The shape of the insulation polyhedron generally corresponds to the shape of the opening in the primary shield. Reactor monitoring instrumentation may be mounted in the corners of the opening in the primary shield between the side walls and the reactor vessel such that insulation is not disposed between the instrumentation and the reactor vessel. 5 figures.

  5. Nuclear reactor having a polyhedral primary shield and removable vessel insulation

    DOE Patents [OSTI]

    Ekeroth, Douglas E. (Delmont, PA); Orr, Richard (Pittsburgh, PA)

    1993-01-01T23:59:59.000Z

    A nuclear reactor is provided having a generally cylindrical reactor vessel disposed within an opening in a primary shield. The opening in the primary shield is defined by a plurality of generally planar side walls forming a generally polyhedral-shaped opening. The reactor vessel is supported within the opening in the primary shield by reactor vessel supports which are in communication and aligned with central portions of some of the side walls. The reactor vessel is connected to the central portions of the reactor vessel supports. A thermal insulation polyhedron formed from a plurality of slidably insertable and removable generally planar insulation panels substantially surrounds at least a portion of the reactor vessel and is disposed between the reactor vessel and the side walls of the primary shield. The shape of the insulation polyhedron generally corresponds to the shape of the opening in the primary shield. Reactor monitoring instrumentation may be mounted in the corners of the opening in the primary shield between the side walls and the reactor vessel such that insulation is not disposed between the instrumentation and the reactor vessel.

  6. A Preliminary Report on Static Analysis of C Code for Nuclear Reactor Protection System

    E-Print Network [OSTI]

    &C system in nuclear power plants consists of various safety and non-safety components. This paper triesA Preliminary Report on Static Analysis of C Code for Nuclear Reactor Protection System Jong-gu Republic of Korea (e-mail: {kirdess, atang34, jbyoo}@konkuk.ac.kr) **Korea Atomic Energy Research Institute

  7. CRAD, Nuclear Safety- Oak Ridge National Laboratory High Flux Isotope Reactor

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Nuclear Safety Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

  8. NA 30 - Deputy Administrator for Naval Reactors | National Nuclear...

    National Nuclear Security Administration (NNSA)

    Us Our Operations Management and Budget Office of Civil Rights Workforce Statistics NA 30 - Deputy Administrator for Naval Reactors NA 30 - Deputy Administrator for...

  9. Safety of Department of Energy-Owned Nuclear Reactors

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1986-09-23T23:59:59.000Z

    To establish reactor safety program requirements assure that the safety of each Department of Energy-owned (DOE-owned) reactor is properly analyzed, evaluated, documented, and approved by DOE; and reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate protection for health and safety and will be in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. Cancels Chap. 6 of DOE O 5480.1A. Paragraphs 7b(3), 7e(3) & 8c canceled by DOE O 5480.23 & canceled by DOE N 251.4 of 9-29-95.

  10. Modularity Approach Modular Pebble Bed Reactor (MPBR)

    E-Print Network [OSTI]

    NED MPBR 1150 MW Combined Heat and Power Station Turbine Hall Boundary Admin Training Control Bldg material limitations #12;4/23/03 MIT NED MPBR Design Elements · Assembly · Self-locating Space

  11. Status of Fuel Development and Manufacturing for Space Nuclear Reactors at BWX Technologies

    SciTech Connect (OSTI)

    Carmack, W.J.; Husser, D.L.; Mohr, T.C.; Richardson, W.C. [BWX Technologies, PO Box 785, Lynchburg, VA 24505-0785 (United States)

    2004-02-04T23:59:59.000Z

    New advanced nuclear space propulsion systems will soon seek a high temperature, stable fuel form. BWX Technologies Inc (BWXT) has a long history of fuel manufacturing. UO2, UCO, and UCx have been fabricated at BWXT for various US and international programs. Recent efforts at BWXT have focused on establishing the manufacturing techniques and analysis capabilities needed to provide a high quality, high power, compact nuclear reactor for use in space nuclear powered missions. To support the production of a space nuclear reactor, uranium nitride has recently been manufactured by BWXT. In addition, analytical chemistry and analysis techniques have been developed to provide verification and qualification of the uranium nitride production process. The fabrication of a space nuclear reactor will require the ability to place an unclad fuel form into a clad structure for assembly into a reactor core configuration. To this end, BWX Technologies has reestablished its capability for machining, GTA welding, and EB welding of refractory metals. Specifically, BWX Technologies has demonstrated GTA welding of niobium flat plate and EB welding of niobium and Nb-1Zr tubing. In performing these demonstration activities, BWX Technologies has established the necessary infrastructure to manufacture UO2, UCx, or UNx fuel, components, and complete reactor assemblies in support of space nuclear programs.

  12. Experimental Results from an Antineutrino Detector for Cooperative Monitoring of Nuclear Reactors

    SciTech Connect (OSTI)

    Bowden, N S; Bernstein, A; Allen, M; Brennan, J S; Cunningham, M; Estrada, J K; Greaves, C R; Hagmann, C; Lund, J; Mengesha, W; Weinbeck, T D; Winant, C D

    2006-09-18T23:59:59.000Z

    Our collaboration has designed, installed, and operated a compact antineutrino detector at a nuclear power station, for the purpose of monitoring the power and plutonium content of the reactor core. This paper focuses on the basic properties and performance of the detector. We describe the site, the reactor source, and the detector, and provide data that clearly show the expected antineutrino signal. Our data and experience demonstrate that it is possible to operate a simple, relatively small, antineutrino detector near a reactor, in a non-intrusive and unattended mode for months to years at a time, from outside the reactor containment, with no disruption of day-to-day operations at the reactor site. This unique real-time cooperative monitoring capability may be of interest for the International Atomic Energy Agency (IAEA) reactor safeguards program and similar regimes.

  13. Safeguards Issues at Nuclear Reactors and Enrichment Plants

    SciTech Connect (OSTI)

    Boyer, Brian D [Los Alamos National Laboratory

    2012-08-15T23:59:59.000Z

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

  14. Hanging core support system for a nuclear reactor. [LMFBR

    DOE Patents [OSTI]

    Burelbach, J.P.; Kann, W.J.; Pan, Y.C.; Saiveau, J.G.; Seidensticker, R.W.

    1984-04-26T23:59:59.000Z

    For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform.

  15. Supplying the nuclear arsenal: Production reactor technology, management, and policy, 1942--1992

    SciTech Connect (OSTI)

    Carlisle, R.P.; Zenzen, J.M.

    1994-01-01T23:59:59.000Z

    This book focuses on the lineage of America`s production reactors, those three at Hanford and their descendants, the reactors behind America`s nuclear weapons. The work will take only occasional sideways glances at the collateral lines of descent, the reactor cousins designed for experimental purposes, ship propulsion, and electric power generation. Over the decades from 1942 through 1992, fourteen American production reactors made enough plutonium to fuel a formidable arsenal of more than twenty thousand weapons. In the last years of that period, planners, nuclear engineers, and managers struggled over designs for the next generation of production reactors. The story of fourteen individual machines and of the planning effort to replace them might appear relatively narrow. Yet these machines lay at the heart of the nation`s nuclear weapons complex. The story of these machines is the story of arming the winning weapon, supplying the nuclear arms race. This book is intended to capture the history of the first fourteen production reactors, and associated design work, in the face of the end of the Cold War.

  16. Advanced Control and Protection system Design Methods for Modular HTGRs

    SciTech Connect (OSTI)

    Ball, Sydney J [ORNL; Wilson Jr, Thomas L [ORNL; Wood, Richard Thomas [ORNL

    2012-06-01T23:59:59.000Z

    The project supported the Nuclear Regulatory Commission (NRC) in identifying and evaluating the regulatory implications concerning the control and protection systems proposed for use in the Department of Energy's (DOE) Next-Generation Nuclear Plant (NGNP). The NGNP, using modular high-temperature gas-cooled reactor (HTGR) technology, is to provide commercial industries with electricity and high-temperature process heat for industrial processes such as hydrogen production. Process heat temperatures range from 700 to 950 C, and for the upper range of these operation temperatures, the modular HTGR is sometimes referred to as the Very High Temperature Reactor or VHTR. Initial NGNP designs are for operation in the lower temperature range. The defining safety characteristic of the modular HTGR is that its primary defense against serious accidents is to be achieved through its inherent properties of the fuel and core. Because of its strong negative temperature coefficient of reactivity and the capability of the fuel to withstand high temperatures, fast-acting active safety systems or prompt operator actions should not be required to prevent significant fuel failure and fission product release. The plant is designed such that its inherent features should provide adequate protection despite operational errors or equipment failure. Figure 1 shows an example modular HTGR layout (prismatic core version), where its inlet coolant enters the reactor vessel at the bottom, traversing up the sides to the top plenum, down-flow through an annular core, and exiting from the lower plenum (hot duct). This research provided NRC staff with (a) insights and knowledge about the control and protection systems for the NGNP and VHTR, (b) information on the technologies/approaches under consideration for use in the reactor and process heat applications, (c) guidelines for the design of highly integrated control rooms, (d) consideration for modeling of control and protection system designs for VHTR, and (e) input for developing the bases for possible new regulatory guidance to assist in the review of an NGNP license application. This NRC project also evaluated reactor and process heat application plant simulation models employed in the protection and control system designs for various plant operational modes and accidents, including providing information about the models themselves, and the appropriateness of the application of the models for control and protection system studies. A companion project for the NRC focused on the potential for new instrumentation that would be unique to modular HTGRs, as compared to light-water reactors (LWRs), due to both the higher temperature ranges and the inherent safety features.

  17. Neural net controlled tag gas sampling system for nuclear reactors

    DOE Patents [OSTI]

    Gross, Kenneth C. (Bolingbrook, IL); Laug, Matthew T. (Idaho Fall, ID); Lambert, John D. B. (Wheaton, IL); Herzog, James P. (Downers Grove, IL)

    1997-01-01T23:59:59.000Z

    A method and system for providing a tag gas identifier to a nuclear fuel rod and analyze escaped tag gas to identify a particular failed nuclear fuel rod. The method and system include disposing a unique tag gas composition into a plenum of a nuclear fuel rod, monitoring gamma ray activity, analyzing gamma ray signals to assess whether a nuclear fuel rod has failed and is emitting tag gas, activating a tag gas sampling and analysis system upon sensing tag gas emission from a failed nuclear rod and evaluating the escaped tag gas to identify the particular failed nuclear fuel rod.

  18. Neural net controlled tag gas sampling system for nuclear reactors

    DOE Patents [OSTI]

    Gross, K.C.; Laug, M.T.; Lambert, J.B.; Herzog, J.P.

    1997-02-11T23:59:59.000Z

    A method and system are disclosed for providing a tag gas identifier to a nuclear fuel rod and analyze escaped tag gas to identify a particular failed nuclear fuel rod. The method and system include disposing a unique tag gas composition into a plenum of a nuclear fuel rod, monitoring gamma ray activity, analyzing gamma ray signals to assess whether a nuclear fuel rod has failed and is emitting tag gas, activating a tag gas sampling and analysis system upon sensing tag gas emission from a failed nuclear rod and evaluating the escaped tag gas to identify the particular failed nuclear fuel rod. 12 figs.

  19. Nuclear reactor with low-level core coolant intake

    DOE Patents [OSTI]

    Challberg, Roy C. (Livermore, CA); Townsend, Harold E. (Campbell, CA)

    1993-01-01T23:59:59.000Z

    A natural-circulation boiling-water reactor has skirts extending downward from control rod guide tubes to about 10 centimeters from the reactor vessel bottom. The skirts define annular channels about control rod drive housings that extend through the reactor vessel bottom. Recirculating water is forced in through the low-level entrances to these channels, sweeping bottom water into the channels in the process. The sweeping action prevents cooler water from accumulating at the bottom. This in turn minimizes thermal shock to bottom-dwelling components as would occur when accumulated cool water is swept away and suddenly replaced by warmer water.

  20. Zirconium-based alloys, nuclear fuel rods and nuclear reactors including such alloys, and related methods

    DOE Patents [OSTI]

    Mariani, Robert Dominick

    2014-09-09T23:59:59.000Z

    Zirconium-based metal alloy compositions comprise zirconium, a first additive in which the permeability of hydrogen decreases with increasing temperatures at least over a temperature range extending from 350.degree. C. to 750.degree. C., and a second additive having a solubility in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. At least one of a solubility of the first additive in the second additive over the temperature range extending from 350.degree. C. to 750.degree. C. and a solubility of the second additive in the first additive over the temperature range extending from 350.degree. C. to 750.degree. C. is higher than the solubility of the second additive in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. Nuclear fuel rods include a cladding material comprising such metal alloy compositions, and nuclear reactors include such fuel rods. Methods are used to fabricate such zirconium-based metal alloy compositions.

  1. Containment building : architecture between the city and advanced nuclear reactors

    E-Print Network [OSTI]

    Pauli, Lisa M

    2011-01-01T23:59:59.000Z

    Since the inception of nuclear energy research, the element thorium (Th) has been considered the superior fuel for nuclear reactions because of its potency, safety, abundance and reduced waste. Cold War agendas broke from ...

  2. Development of Technical Nuclear Forensics for Spent Research Reactor Fuel

    E-Print Network [OSTI]

    Sternat, Matthew Ryan 1982-

    2012-11-20T23:59:59.000Z

    , an inverse analysis was developed to re-construct the burnup, initial uranium isotopic compositions, and cooling time of a research reactor spent fuel sample. A convergence acceleration technique was used that consisted of an analytical calculation to predict...

  3. The Neutrino Mass Hierarchy from Nuclear Reactor Experiments

    E-Print Network [OSTI]

    Emilio Ciuffoli; Jarah Evslin; Xinmin Zhang

    2013-08-14T23:59:59.000Z

    10 years from now reactor neutrino experiments will attempt to determine which neutrino mass eigenstate is the most massive. In this letter we present the results of more than seven million detailed simulations of such experiments, studying the dependence of the probability of successfully determining the mass hierarchy upon the analysis method, the neutrino mass matrix parameters, reactor flux models, geoneutrinos and, in particular, combinations of baselines. We show that a recently reported spurious dependence of the data analysis upon the high energy tail of the reactor spectrum can be removed by using a weighted Fourier transform. We determine the optimal baselines and corresponding detector locations. For most values of the CP-violating, leptonic Dirac phase delta, a degeneracy prevents NOvA and T2K from determining either delta or the hierarchy. We determine the confidence with which a reactor experiment can determine the hierarchy, breaking the degeneracy.

  4. Argonne partners with industry on nuclear reactor work | Argonne...

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

    and the University of Pittsburgh to develop thermo-acoustic sensors for sodium-cooled fast reactors. The awards are part of the Obama administration's all-of-the-above energy...

  5. Dual annular rotating [open quotes]windowed[close quotes] nuclear reflector reactor control system

    DOE Patents [OSTI]

    Jacox, M.G.; Drexler, R.L.; Hunt, R.N.M.; Lake, J.A.

    1994-03-29T23:59:59.000Z

    A nuclear reactor control system is provided in a nuclear reactor having a core operating in the fast neutron energy spectrum where criticality control is achieved by neutron leakage. The control system includes dual annular, rotatable reflector rings. There are two reflector rings: an inner reflector ring and an outer reflector ring. The reflectors are concentrically assembled, surround the reactor core, and each reflector ring includes a plurality of openings. The openings in each ring are capable of being aligned or non-aligned with each other. Independent driving means for each of the annular reflector rings is provided so that reactor criticality can be initiated and controlled by rotation of either reflector ring such that the extent of alignment of the openings in each ring controls the reflection of neutrons from the core. 4 figures.

  6. Passive cooling system for top entry liquid metal cooled nuclear reactors

    DOE Patents [OSTI]

    Boardman, Charles E. (Saratoga, CA); Hunsbedt, Anstein (Los Gatos, CA); Hui, Marvin M. (Cupertino, CA)

    1992-01-01T23:59:59.000Z

    A liquid metal cooled nuclear fission reactor plant having a top entry loop joined satellite assembly with a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during shutdown, or heat produced during a mishap. This satellite type reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary cooling system when rendered inoperative.

  7. Blue Ribbon Commission, Yucca Mountain Closure, Court Actions - Future of Decommissioned Reactors, Operating Reactors and Nuclear Power - 13249

    SciTech Connect (OSTI)

    Devgun, Jas S. [Nuclear Power Technologies, Sargent and Lundy LLC1, Chicago, IL (United States)] [Nuclear Power Technologies, Sargent and Lundy LLC1, Chicago, IL (United States)

    2013-07-01T23:59:59.000Z

    Issues related to back-end of the nuclear fuel cycle continue to be difficult for the commercial nuclear power industry and for the decision makers at the national and international level. In the US, the 1982 NWPA required DOE to develop geological repositories for SNF and HLW but in spite of extensive site characterization efforts and over ten billion dollars spent, a repository opening is nowhere in sight. There has been constant litigation against the DOE by the nuclear utilities for breach of the 'standard contract' they signed with the DOE under the NWPA. The SNF inventory continues to rise both in the US and globally and the nuclear industry has turned to dry storage facilities at reactor locations. In US, the Blue Ribbon Commission on America's Nuclear Future issued its report in January 2012 and among other items, it recommends a new, consent-based approach to siting of facilities, prompt efforts to develop one or more geologic disposal facilities, and prompt efforts to develop one or more consolidated storage facilities. In addition, the March 2011 Fukushima Daiichi accident had a severe impact on the future growth of nuclear power. The nuclear industry is focusing on mitigation strategies for beyond design basis events and in the US, the industry is in the process of implementing the recommendations from NRC's Near Term Task Force. (authors)

  8. Detecting a Nuclear Fission Reactor at the Center of the Earth

    E-Print Network [OSTI]

    R. S. Raghavan

    2002-08-24T23:59:59.000Z

    A natural nuclear fission reactor with a power output of 3- 10 terawatt at the center of the earth has been proposed as the energy source of the earth's magnetic field. The proposal can be directly tested by a massive liquid scintillation detector that can detect the signature spectrum of antineutrinos from the geo-reactor as well as the direction of the antineutrino source. Such detectors are now in operation or under construction in Japan/Europe. However, the clarity of both types of measurements may be limited by background from antineutrinos from surface power reactors. Future U. S. detectors, relatively more remote from power reactors, may be more suitable for achieving unambiguous spectral and directional evidence for a 3TW geo-reactor.

  9. Reactor Physics and Criticality Benchmark Evaluations for Advanced Nuclear Fuel - Final Technical Report

    SciTech Connect (OSTI)

    William Anderson; James Tulenko; Bradley Rearden; Gary Harms

    2008-09-11T23:59:59.000Z

    The nuclear industry interest in advanced fuel and reactor design often drives towards fuel with uranium enrichments greater than 5 wt% 235U. Unfortunately, little data exists, in the form of reactor physics and criticality benchmarks, for uranium enrichments ranging between 5 and 10 wt% 235U. The primary purpose of this project is to provide benchmarks for fuel similar to what may be required for advanced light water reactors (LWRs). These experiments will ultimately provide additional information for application to the criticality-safety bases for commercial fuel facilities handling greater than 5 wt% 235U fuel.

  10. Utilizing a Russian space nuclear reactor for a United States space mission: Systems integration issues

    SciTech Connect (OSTI)

    Reynolds, E.; Schaefer, E. [Johns Hopkins Univ., Laurel, MD (United States). Applied Physics Lab.; Polansky, G.; Lacy, J. [Phillips Lab., Albuquerque, NM (United States); Bocharov, A. [GDBMB, St. Petersburg (Russian Federation)

    1993-09-30T23:59:59.000Z

    The Nuclear Electric Propulsion Space Test Program (NEPSTP) has developed a cooperative relationship with several institutes of the former Soviet Union to evaluate Russian space hardware on a US spacecraft One component is the Topaz II Nuclear Power System; a built and flight qualified nuclear reactor that has yet to be tested in space. The access to the Topaz II reactor provides the NEPSTP with a rare opportunity; to conduct an early flight demonstration of nuclear electric propulsion at a relatively low cost. This opportunity, however, is not without challenges. Topaz II was designed to be compatible with Russian spacecraft and launch vehicles. It was manufactured and flight qualified by Russian techniques and standards and conforms to safety requirements of the former Soviet Union, not the United States. As it is desired to make minimal modifications to the Topaz II, integrating the reactor system with a United States spacecraft and launch vehicle presents an engineering challenge. This paper documents the lessons teamed regarding the integration of reactor based spacecraft and also some insight about integrating Russian hardware. It examines the planned integration flow along with specific reactor requirements that affect the spacecraft integration including American-Russian space system compatibility.

  11. Three dimensional neutronics calculations for the TAMU Nuclear Science Center Triga reactor using BOLD VENTURE

    E-Print Network [OSTI]

    Yupari, Ricardo

    1985-01-01T23:59:59.000Z

    using the computer code BOLD VENTURE. BOLD VENTURE is a modular type computational system which was developed by Oak Ridge National Laborator y for nuclear r eactor analysis. VENTURE is the neutronics module which implements a f inite...-dif'ference solution of the multigroup diffusion equations in 3-D geometry. The computer code was loaded and compiled on the TAMU AMDAHL computer, and sample pr oblems were run in or der to test the performance of each of the computational modules. The effective...

  12. Use of MCNP for characterization of reactor vessel internals waste from decommissioned nuclear reactors

    SciTech Connect (OSTI)

    Love, E.F.; Pauley, K.A.; Reid, B.D.

    1995-09-01T23:59:59.000Z

    This study describes the use of the Monte Carlo Neutron-Photon (MCNP) code for determining activation levels of irradiated reactor vessel internals hardware. The purpose of the analysis is to produce data for the Department of Energy`s Greater-Than-Class C Low-Level Radioactive Waste Program. An MCNP model was developed to analyze the Yankee Rowe reactor facility. The model incorporates reactor geometry, material compositions, and operating history data acquired from Yankee Atomic Electric Company. In addition to the base activation analysis, parametric studies were performed to determine the sensitivity of activation to specific parameters. A component sampling plan was also developed to validate the model results, although the plan was not implemented. The calculations for the Yankee Rowe reactor predict that only the core baffle and the core support plates will be activated to levels above the Class C limits. The parametric calculations show, however, that the large uncertainties in the material compositions could cause errors in the estimates that could also increase the estimated activation level of the core barrel to above the Class C limits. Extrapolation of the results to other reactor facilities indicates that in addition to the baffle and support plates, core barrels may also be activated to above Class C limits; however the classification will depend on the specific operating conditions of the reactor and the specific material compositions of the metal, as well as the use of allowable concentration averaging practices in packaging and classifying the waste.

  13. MANAGEMENT OF RESEARCH AND TEST REACTOR ALUMINUM SPENT NUCLEAR FUEL - A TECHNOLOGY ASSESSMENT

    SciTech Connect (OSTI)

    Vinson, D.

    2010-07-11T23:59:59.000Z

    The Department of Energy's Environmental Management (DOE-EM) Program is responsible for the receipt and storage of aluminum research reactor spent nuclear fuel or used fuel until ultimate disposition. Aluminum research reactor used fuel is currently being stored or is anticipated to be returned to the U.S. and stored at DOE-EM storage facilities at the Savannah River Site and the Idaho Nuclear Technology and Engineering Center. This paper assesses the technologies and the options for safe transportation/receipt and interim storage of aluminum research reactor spent fuel and reviews the comprehensive strategy for its management. The U.S. Department of Energy uses the Appendix A, Spent Nuclear Fuel Acceptance Criteria, to identify the physical, chemical, and isotopic characteristics of spent nuclear fuel to be returned to the United States under the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The fuel is further evaluated for acceptance through assessments of the fuel at the foreign sites that include corrosion damage and handleability. Transport involves use of commercial shipping casks with defined leakage rates that can provide containment of the fuel, some of which are breached. Options for safe storage include wet storage and dry storage. Both options must fully address potential degradation of the aluminum during the storage period. This paper focuses on the various options for safe transport and storage with respect to technology maturity and application.

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

    E-Print Network [OSTI]

    Terrani, Kurt Amir

    2010-01-01T23:59:59.000Z

    1.2.1 PWRs . . . . . . . . . . . . . . . . . . . . 1.2.2Actinides Multi-Recycling in PWR Using Hydride Fuels. InRecycling in Hydride Fueled PWR Cores. Nuclear Engineering

  15. Composite Materials under Extreme Radiation and Temperature Environments of the Next Generation Nuclear Reactors

    SciTech Connect (OSTI)

    Simos, N.

    2011-05-01T23:59:59.000Z

    In the nuclear energy renaissance, driven by fission reactor concepts utilizing very high temperatures and fast neutron spectra, materials with enhanced performance that exceeds are expected to play a central role. With the operating temperatures of the Generation III reactors bringing the classical reactor materials close to their performance limits there is an urgent need to develop and qualify new alloys and composites. Efforts have been focused on the intricate relations and the high demands placed on materials at the anticipated extreme states within the next generation fusion and fission reactors which combine high radiation fluxes, elevated temperatures and aggressive environments. While nuclear reactors have been in operation for several decades, the structural materials associated with the next generation options need to endure much higher temperatures (1200 C), higher neutron doses (tens of displacements per atom, dpa), and extremely corrosive environments, which are beyond the experience on materials accumulated to-date. The most important consideration is the performance and reliability of structural materials for both in-core and out-of-core functions. While there exists a great body of nuclear materials research and operating experience/performance from fission reactors where epithermal and thermal neutrons interact with materials and alter their physio-mechanical properties, a process that is well understood by now, there are no operating or even experimental facilities that will facilitate the extreme conditions of flux and temperature anticipated and thus provide insights into the behaviour of these well understood materials. Materials, however, still need to be developed and their interaction and damage potential or lifetime to be quantified for the next generation nuclear energy. Based on material development advances, composites, and in particular ceramic composites, seem to inherently possess properties suitable for key functions within the operating envelope of both fission and fusion reactors. In advanced fission reactors composite materials are being designed in an effort to extend the life and improve the reliability of fuel rod cladding as well as structural materials. Composites are being considered for use as core internals in the next generation of gas-cooled reactors. Further, next-generation plasma-fusion reactors, such as the International Thermonuclear Experimental Reactor (ITER) will rely on the capabilities of advanced composites to safely withstand extremely high neutron fluxes while providing superior thermal shock resistance.

  16. DETERMINING THE EFFECTS OF RADIATION ON AGING CONCRETE STRUCTURES OF NUCLEAR REACTORS

    SciTech Connect (OSTI)

    Serrato, M.

    2010-01-29T23:59:59.000Z

    The U.S. Department of Energy Office of Environmental Management (DOE-EM) is responsible for the Decontamination and Decommissioning (D&D) of nuclear facilities throughout the DOE Complex. Some of these facilities will be completely dismantled, while others will be partially dismantled and the remaining structure will be stabilized with cementitious fill materials. The latter is a process known as In-Situ Decommissioning (ISD). The ISD decision process requires a detailed understanding of the existing facility conditions, and operational history. System information and material properties are need for aged nuclear facilities. This literature review investigated the properties of aged concrete structures affected by radiation. In particular, this review addresses the Savannah River Site (SRS) isotope production nuclear reactors. The concrete in the reactors at SRS was not seriously damaged by the levels of radiation exposure. Loss of composite compressive strength was the most common effect of radiation induced damage documented at nuclear power plants.

  17. Nuclear breeder reactor fuel element with silicon carbide getter

    DOE Patents [OSTI]

    Christiansen, David W. (Kennewick, WA); Karnesky, Richard A. (Richland, WA)

    1987-01-01T23:59:59.000Z

    An improved cesium getter 28 is provided in a breeder reactor fuel element or pin in the form of an extended surface area, low density element formed in one embodiment as a helically wound foil 30 located with silicon carbide, and located at the upper end of the fertile material upper blanket 20.

  18. Nuclear Reactor/Hydrogen Process Interface Including the HyPEP Model

    SciTech Connect (OSTI)

    Steven R. Sherman

    2007-05-01T23:59:59.000Z

    The Nuclear Reactor/Hydrogen Plant interface is the intermediate heat transport loop that will connect a very high temperature gas-cooled nuclear reactor (VHTR) to a thermochemical, high-temperature electrolysis, or hybrid hydrogen production plant. A prototype plant called the Next Generation Nuclear Plant (NGNP) is planned for construction and operation at the Idaho National Laboratory in the 2018-2021 timeframe, and will involve a VHTR, a high-temperature interface, and a hydrogen production plant. The interface is responsible for transporting high-temperature thermal energy from the nuclear reactor to the hydrogen production plant while protecting the nuclear plant from operational disturbances at the hydrogen plant. Development of the interface is occurring under the DOE Nuclear Hydrogen Initiative (NHI) and involves the study, design, and development of high-temperature heat exchangers, heat transport systems, materials, safety, and integrated system models. Research and development work on the system interface began in 2004 and is expected to continue at least until the start of construction of an engineering-scale demonstration plant.

  19. Terracentric Nuclear Fission Reactor: Background, Basis, Feasibility, Structure, Evidence, and Geophysical Implications

    E-Print Network [OSTI]

    J. Marvin Herndon

    2013-12-31T23:59:59.000Z

    The background, basis, feasibility, structure, evidence, and geophysical implications of a naturally occurring Terracentric nuclear fission georeactor are reviewed. For a nuclear fission reactor to exist at the center of the Earth, all of the following conditions must be met: (1) There must originally have been a substantial quantity of uranium within Earth's core; (2) There must be a natural mechanism for concentrating the uranium; (3) The isotopic composition of the uranium at the onset of fission must be appropriate to sustain a nuclear fission chain reaction; (4) The reactor must be able to breed a sufficient quantity of fissile nuclides to permit operation over the lifetime of Earth to the present; (5) There must be a natural mechanism for the removal of fission products; (6) There must be a natural mechanism for removing heat from the reactor; (7) There must be a natural mechanism to regulate reactor power level, and; (8) The location of the reactor or must be such as to provide containment and prevent meltdown. Herndon's georeactor alone is shown to meet those conditions. Georeactor existence evidence based upon helium measurements and upon antineutrino measurements is described. Geophysical implications discussed include georeactor origin of the geomagnetic field, geomagnetic reversals from intense solar outbursts and severe Earth trauma, as well as georeactor heat contributions to global dynamics.

  20. Terracentric Nuclear Fission Reactor: Background, Basis, Feasibility, Structure, Evidence, and Geophysical Implications

    E-Print Network [OSTI]

    Herndon, J Marvin

    2013-01-01T23:59:59.000Z

    The background, basis, feasibility, structure, evidence, and geophysical implications of a naturally occurring Terracentric nuclear fission georeactor are reviewed. For a nuclear fission reactor to exist at the center of the Earth, all of the following conditions must be met: (1) There must originally have been a substantial quantity of uranium within Earth's core; (2) There must be a natural mechanism for concentrating the uranium; (3) The isotopic composition of the uranium at the onset of fission must be appropriate to sustain a nuclear fission chain reaction; (4) The reactor must be able to breed a sufficient quantity of fissile nuclides to permit operation over the lifetime of Earth to the present; (5) There must be a natural mechanism for the removal of fission products; (6) There must be a natural mechanism for removing heat from the reactor; (7) There must be a natural mechanism to regulate reactor power level, and; (8) The location of the reactor or must be such as to provide containment and prevent ...

  1. International Nuclear Energy Policy and Cooperation | Department...

    Energy Savers [EERE]

    encompasses technologies related to small modular reactors (SMRs), sodium-cooled fast reactors, light water reactor accident-tolerant fuels, actinide separations and waste...

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

    SciTech Connect (OSTI)

    M. J. Russell

    2006-06-01T23:59:59.000Z

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

  3. Advanced neutron irradiation system using Texas A&M University Nuclear Science Center Reactor

    E-Print Network [OSTI]

    Jang, Si Young

    2005-11-01T23:59:59.000Z

    was installed in the irradiation cell of the Texas A&M University Nuclear Science Center Reactor (NSCR). By increasing the thickness of the lead-bismuth alloy, the neutron spectra were shifted into lower energies by the scattering interactions of fast...

  4. EU in push for support on nuclear fusion reactor September 26, 2004

    E-Print Network [OSTI]

    to obtain power through nuclear fusion, a clean energy source. But views are split on where the ITER reactor, estimated to cost around $10 billion over 30 years, should be based. While the EU, backed by China yesterday. Diplomats said the US resistance seemed to be political rather than based on scientific grounds

  5. Investigation of Neutrino Properties in Experiments at Nuclear Reactors: Present Status and Prospects

    E-Print Network [OSTI]

    L. A. Mikaelyan

    2002-10-07T23:59:59.000Z

    This paper was submitted in Russian edition of Journal Physics of Atomic Nuclei in 2001. The present status of experiments that are being performed at nuclear reactors in order to seek the neutrino masses, mixing, and magnetic moments, whose discovery would be a signal of the existence of physics beyond the Standard Model, is considered, along with their future prospects.

  6. Spring design for use in the core of a nuclear reactor

    DOE Patents [OSTI]

    Willard, Jr., H. James (Bethel Park, PA)

    1993-01-01T23:59:59.000Z

    A spring design particularly suitable for use in the core of a nuclear reactor includes one surface having a first material oriented in a longitudinal direction, and another surface having a second material oriented in a transverse direction. The respective surfaces exhibit different amounts of irraditation induced strain.

  7. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    SciTech Connect (OSTI)

    Not Available

    1993-11-01T23:59:59.000Z

    This document contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE non-reactor nuclear facilities. Adherence to these guidelines will provide consistency and uniformity in criticality safety evaluations (CSEs) across the complex and will document compliance with the requirements of DOE Order 5480.24.

  8. Knowledges and abilities catalog for nuclear power plant operators: pressurized water reactors

    SciTech Connect (OSTI)

    Not Available

    1985-07-01T23:59:59.000Z

    This document catalogs roughly 5300 knowledges and abilities of reactor operators and senior reactor operators. It results from a reanalysis of much larger job-task analysis data base compiled by the Institute of Nuclear Power Operations (INPO). Knowledges and abilities are cataloged for 45 major power plant systems and 38 emergency evolutions, grouped according to 11 fundamental safety functions (e.g., reactivity control and reactor coolant system inventory control). With appropriate sampling from this catalog, operator licensing examinations having content validity can be developed. A structured sampling procedure for this catalog is under development by the Nuclear Regulatory Commission (NRC) and will be published as a companion document, ''Examiners' Handbook for Developing Operator Licensing Examinations'' (NUREG-1121). The examinations developed by using the catalog and handbook will cover those topics listed under Title 10, Code of Federal Regulations, Part 55.

  9. FLOWSHEET EVALUATION FOR THE DISSOLVING AND NEUTRALIZATION OF SODIUM REACTOR EXPERIMENT USED NUCLEAR FUEL

    SciTech Connect (OSTI)

    Daniel, W. E.; Hansen, E. K.; Shehee, T. C.

    2012-10-30T23:59:59.000Z

    This report includes the literature review, hydrogen off-gas calculations, and hydrogen generation tests to determine that H-Canyon can safely dissolve the Sodium Reactor Experiment (SRE; thorium fuel), Ford Nuclear Reactor (FNR; aluminum alloy fuel), and Denmark Reactor (DR-3; silicide fuel, aluminum alloy fuel, and aluminum oxide fuel) assemblies in the L-Bundles with respect to the hydrogen levels in the projected peak off-gas rates. This is provided that the number of L-Bundles charged to the dissolver is controlled. Examination of SRE dissolution for potential issues has aided in predicting the optimal batching scenario. The calculations detailed in this report demonstrate that the FNR, SRE, and DR-3 used nuclear fuel (UNF) are bounded by MURR UNF and may be charged using the controls outlined for MURR dissolution in a prior report.

  10. International Conference on the Physics of Reactors "Nuclear Power: A Sustainable Resource" Casino-Kursaal Conference Center, Interlaken, Switzerland, September 14-19, 2008

    E-Print Network [OSTI]

    Boyer, Edmond

    International Conference on the Physics of Reactors "Nuclear Power: A Sustainable Resource" Casino International Forum for the new nuclear energy systems, we have developed a new concept of molten salt reactor Products which poison the core can be extracted without stopping reactor operation; nuclear waste

  11. Japan hopes for rapid accord with EU on nuclear reactor Japanese Prime Minister Junichiro Koizumi said Monday that he hopes to resolve a dispute with the

    E-Print Network [OSTI]

    Japan hopes for rapid accord with EU on nuclear reactor 02/05/2005 Japanese Prime Minister nuclear reactor as soon as possible. Koizumi discussed the row over plans for the International Terms and Conditions 5/2/05 8:59 AMEUbusiness - Japan hopes for rapid accord with EU on nuclear reactor

  12. For more information, contact Michele Boyd at mboyd@psr.org. Updated July 13, 2009. Existing Subsidies and Incentives for New Nuclear Reactors

    E-Print Network [OSTI]

    Laughlin, Robert B.

    Subsidies and Incentives for New Nuclear Reactors Research and Development · Generation IV program in construction and operation licensing for 6 new reactors, including delays due to the Nuclear Regulatory-hour paid by ratepayers receiving electricity from nuclear reactors to pay for a geologic repository

  13. International Topical Meeting on Nuclear Reactor Thermal hydraulics, NURETH-15 NURETH15-599 Pisa, Italy, May 12-17, 2013

    E-Print Network [OSTI]

    Haviland, David

    The 15th International Topical Meeting on Nuclear Reactor Thermal hydraulics, NURETH-15 NURETH15 of steam explosion shock wave propagation. #12;The 15th International Topical Meeting on Nuclear Reactor to note that stratified melt-coolant configuration can be formed in several scenarios of nuclear reactor

  14. Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

    SciTech Connect (OSTI)

    Shropshire, D.E.; Herring, J.S.

    2004-10-03T23:59:59.000Z

    The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrain all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim storage, packaging, transportation, waste forms, waste treatment, decontamination and decommissioning issues; and low-level waste (LLW) and high-level waste (HLW) disposal.

  15. Energy Secretary to Visit Georgia Nuclear Reactor Site and Tennessee...

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

    D.C. - U.S. Secretary of Energy Secretary Steven Chu will visit the Vogtle nuclear power plant in Waynesboro, Georgia, and Oak Ridge National Laboratory on Wednesday,...

  16. Nuclear processes in magnetic fusion reactors with polarized fuel

    E-Print Network [OSTI]

    Michail P. Rekalo; Egle Tomasi-Gustafsson

    2000-10-16T23:59:59.000Z

    We consider the processes $d +d \\to n +{^3He}$, $d +{^3He} \\to p +{^4He}$, $d +{^3H} \\to n +{^4He}$, ${^3He} +{^3He}\\to p+p +{^4He}$, ${^3H} +{^3He}\\to d +{^4He}$, with particular attention for applications in fusion reactors. After a model independent parametrization of the spin structure of the matrix elements for these processes at thermal colliding energies, in terms of partial amplitudes, we study polarization phenomena in the framework of a formalism of helicity amplitudes. The strong angular dependence of the final nuclei and of the polarization observables on the polarizations of the fuel components can be helpful in the design of the reactor shielding, blanket arrangement etc..We analyze also the angular dependence of the neutron polarization for the processes $\\vec d +\\vec d \\to n +{^3He}$ and $\\vec d +\\vec {^3H} \\to n +{^4He}$.

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

    DOE Patents [OSTI]

    Forsberg, Charles W. (Oak Ridge, TN)

    1991-01-01T23:59:59.000Z

    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.

  18. Reactivity worth measurements at the IPEN/MB-01 nuclear reactor

    SciTech Connect (OSTI)

    Pinto, Leticia Negrao; Santos, Adimir dos [Nuclear Engineering Center, Nuclear and Energy Research Institute- IPEN/CNEN-SP (Brazil)

    2013-05-06T23:59:59.000Z

    Researches that aim to improve the performance of neutron transport codes and quality of nuclear cross section databases are very important to increase the accuracy of simulations and the quality of the analysis and prediction of phenomena in the nuclear field. In this context, relevant experimental data such as reactivity worth measurements are needed. The objective of this work was to perform a series of experiments of reactivity worth measurements, using a digital reactivity meter developed at IPEN. The experiments employed small metallic and ceramic samples inserted in the central region of the core of the experimental IPEN/MB-01 reactor. The theoretical analysis was performed by the MCNP-5 reactor physics code, developed and maintained by Los Alamos National Laboratory, and the ENDF/B-VII.0 nuclear data library.

  19. Fresh and Spent Nuclear Fuel Repatriation from the IRT-2000 Research Reactor Facility, Sofia, Bulgaria

    SciTech Connect (OSTI)

    K. J. Allen; T. G. Apostolov; I. S. Dimitrov

    2009-03-01T23:59:59.000Z

    The IRT 2000 research reactor, operated by the Bulgarian Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped all of their Russian-origin nuclear fuel from the Republic of Bulgaria to the Russian Federation beginning in 2003 and completing in 2008. These fresh and spent fuel shipments removed all highly enriched uranium (HEU) from Bulgaria. The fresh fuel was shipped by air in December 2003 using trucks and a commercial cargo aircraft. One combined spent fuel shipment of HEU and low enriched uranium (LEU) was completed in July 2008 using high capacity VPVR/M casks transported by truck, barge, and rail. The HEU shipments were assisted by the Russian Research Reactor Fuel Return Program (RRRFR) and the LEU spent fuel shipment was funded by Bulgaria. This report describes the work, approvals, organizations, equipment, and agreements required to complete these shipments and concludes with several major lessons learned.

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

    SciTech Connect (OSTI)

    S. Blaine Grover

    2009-09-01T23:59:59.000Z

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

  1. Evaluation of a Business Case for Safeguards by Design in Nuclear Power Reactors

    SciTech Connect (OSTI)

    Wood, Thomas W.; Seward, Amy M.; Lewis, Valerie A.; Gitau, Ernest TN; Zentner, Michael D.

    2012-12-01T23:59:59.000Z

    Safeguards by Design (SbD) is a well-known paradigm for consideration and incorporation of safeguards approaches and associated design features early in the nuclear facility development process. This paradigm has been developed as part of the Next Generation Safeguards Initiative (NGSI), and has been accepted as beneficial in many discussions and papers on NGSI or specific technologies under development within NGSI. The Office of Nuclear Safeguards and Security funded the Pacific Northwest National Laboratory to examine the business case justification of SbD for nuclear power reactors. Ultimately, the implementation of SbD will rely on the designers of nuclear facilities. Therefore, it is important to assess the incentives which will lead designers to adopt SbD as a standard practice for nuclear facility design. This report details the extent to which designers will have compelling economic incentives to adopt SbD.

  2. Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris

    DOE Patents [OSTI]

    Gabor, John D. (Western Springs, IL); Cassulo, John C. (Stickney, IL); Pedersen, Dean R. (Naperville, IL); Baker, Jr., Louis (Downers Grove, IL)

    1986-01-01T23:59:59.000Z

    The invention teaches safety apparatus that can be included in a nuclear reactor, either when newly fabricated or as a retrofit add-on, that will minimize proliferation of structural damage to the reactor in the event the reactor is experiencing an overheating malfunction whereby radioactive nuclear debris might break away from and be discharged from the reactor core. The invention provides a porous bed or sublayer on the lower surface of the reactor containment vessel so that the debris falls on and piles up on the bed. Vapor release elements upstand from the bed in some laterally spaced array. Thus should the high heat flux of the debris interior vaporize the coolant at that location, the vaporized coolant can be vented downwardly to and laterally through the bed to the vapor release elements and in turn via the release elements upwardly through the debris. This minimizes the pressure buildup in the debris and allows for continuing infiltration of the liquid coolant into the debris interior.

  3. 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

    2010-01-01T23:59:59.000Z

    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...

  4. Environmental Assessment of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Not Available

    1994-04-01T23:59:59.000Z

    The Department of Energy has completed the Environmental Assessment (EA) of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel and issued a Finding of No Significant Impact (FONSI) for the proposed action. The EA and FONSI are enclosed for your information. The Department has decided to accept a limited number of spent nuclear fuel elements (409 elements) containing uranium that was enriched in the United States from eight research reactors in Austria, Denmark, Germany, Greece, the Netherlands, Sweden, and Switzerland. This action is necessary to maintain the viability of a major US nuclear weapons nonproliferation program to limit or eliminate the use of highly enriched uranium in civil programs. The purpose of the EA is to maintain the cooperation of the foreign research reactor operators with the nonproliferation program while a more extensive Environmental Impact Statement (EIS) is prepared on a proposed broader policy involving the acceptance of up to 15,000 foreign research reactor spent fuel elements over a 10 to 15 year period. Based on an evaluation of transport by commercial container liner or chartered vessel, five eastern seaboard ports, and truck and train modes of transporting the spent fuel overland to the Savannah River Sits, the Department has concluded that no significant impact would result from any combination of port and made of transport. In addition, no significant impacts were found from interim storage of spent fuel at the Savannah River Site.

  5. Monitoring system for a liquid-cooled nuclear fission reactor. [PWR

    DOE Patents [OSTI]

    DeVolpi, A.

    1984-07-20T23:59:59.000Z

    The invention provides improved means for detecting the water levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting the density of the water in these regions. The invention utilizes a plurality of exterior gamma radiation detectors and a collimator technique operable to sense separate regions of the reactor vessel to give respectively, unique signals for these regions, whereby comparative analysis of these signals can be used to advise of the presence and density of cooling water in the vessel.

  6. Acoustic emission monitoring of hot functional testing: Watts Bar Unit 1 Nuclear Reactor

    SciTech Connect (OSTI)

    Hutton, P.H.; Dawson, J.F.; Friesel, M.A.; Harris, J.C.; Pappas, R.A.

    1984-06-01T23:59:59.000Z

    Acoustic emission (AE) monitoring of selected pressure boundary areas at TVA's Watts Bar, Unit 1 Nuclear Power Plant during hot functional preservice testing is described in this report. The report deals with background, methodology, and results. The work discussed here is a major milestone in a program supported by NRC to develop and demonstrate application of AE monitoring for continuous surveillance of reactor pressure boundaries to detect and evaluate growing flaws. The subject work demonstrated that anticipated problem areas can be overcome. Work is continuing toward AE monitoring during reactor operation.

  7. Can Next-Generation Reactors Power a Safe Nuclear Futur By Clay Dillow Posted 03.17.2011 at 12:18 pm

    E-Print Network [OSTI]

    Danon, Yaron

    Can Next-Generation Reactors Power a Safe Nuclear Futur By Clay Dillow Posted 03.17.2011 at 12 of nuclear reactors are designed to prevent exactly what we old Fukushima Daiichi plant. Which is good the world rush to reconsider their nuclear plans, nuclear experts look toward a future of smaller, safer

  8. Nuclear reactor melt-retention structure to mitigate direct containment heating

    DOE Patents [OSTI]

    Tutu, Narinder K. (Manorville, NY); Ginsberg, Theodore (East Setauket, NY); Klages, John R. (Mattituck, NY)

    1991-01-01T23:59:59.000Z

    A light water nuclear reactor melt-retention structure to mitigate the extent of direct containment heating of the reactor containment building. The structure includes a retention chamber for retaining molten core material away from the upper regions of the reactor containment building when a severe accident causes the bottom of the pressure vessel of the reactor to fail and discharge such molten material under high pressure through the reactor cavity into the retention chamber. In combination with the melt-retention chamber there is provided a passageway that includes molten core droplet deflector vanes and has gas vent means in its upper surface, which means are operable to deflect molten core droplets into the retention chamber while allowing high pressure steam and gases to be vented into the upper regions of the containment building. A plurality of platforms are mounted within the passageway and the melt-retention structure to direct the flow of molten core material and help retain it within the melt-retention chamber. In addition, ribs are mounted at spaced positions on the floor of the melt-retention chamber, and grid means are positioned at the entrance side of the retention chamber. The grid means develop gas back pressure that helps separate the molten core droplets from discharged high pressure steam and gases, thereby forcing the steam and gases to vent into the upper regions of the reactor containment building.

  9. Considerations Associated with Reactor Technology Selection for the Next Generation Nuclear Plant Project

    SciTech Connect (OSTI)

    L.E. Demick

    2010-09-01T23:59:59.000Z

    At the inception of the Next Generation Nuclear Plant Project and during predecessor activities, alternative reactor technologies have been evaluated to determine the technology that best fulfills the functional and performance requirements of the targeted energy applications and market. Unlike the case of electric power generation where the reactor performance is primarily expressed in terms of economics, the targeted energy applications involve industrial applications that have specific needs in terms of acceptable heat transport fluids and the associated thermodynamic conditions. Hence, to be of interest to these industrial energy applications, the alternative reactor technologies are weighed in terms of the reactor coolant/heat transport fluid, achievable reactor outlet temperature, and practicality of operations to achieve the very high reliability demands associated with the petrochemical, petroleum, metals and related industries. These evaluations have concluded that the high temperature gas-cooled reactor (HTGR) can uniquely provide the required ranges of energy needs for these target applications, do so with promising economics, and can be commercialized with reasonable development risk in the time frames of current industry interest – i.e., within the next 10-15 years.

  10. TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

    SciTech Connect (OSTI)

    Ryan, B.C.

    1997-05-01T23:59:59.000Z

    This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.

  11. Nuclear reactor heat transport system component low friction support system

    DOE Patents [OSTI]

    Wade, Elman E. (Ruffs Dale, PA)

    1980-01-01T23:59:59.000Z

    A support column for a heavy component of a liquid metal fast breeder reactor heat transport system which will deflect when the pipes leading coolant to and from the heavy component expand or contract due to temperature changes includes a vertically disposed pipe, the pipe being connected to the heavy component by two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles and the pipe being supported through two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles, each of the cylindrical surfaces bearing on a flat and horizontal surface.

  12. Method for removing cesium from a nuclear reactor coolant

    DOE Patents [OSTI]

    Colburn, R.P.

    1983-08-10T23:59:59.000Z

    A method of and system for removing cesium from a liquid metal reactor coolant including a carbon packing trap in the primary coolant system for absorbing a major portion of the radioactive cesium from the coolant flowing therethrough at a reduced temperature. A regeneration subloop system having a secondary carbon packing trap is selectively connected to the primary system for isolating the main trap therefrom and connecting it to the regeneration system. Increasing the temperature of the sodium flowing through the primary trap diffuses a portion of the cesium inventory thereof further into the carbon matrix while simultaneously redispersing a portion into the regeneration system for absorption at a reduced temperature by the secondary trap.

  13. Device for actuating a nuclear reactor control rod

    SciTech Connect (OSTI)

    Etienne, C.; Fardouet, C.; L'abbe, G.M.

    1982-02-09T23:59:59.000Z

    A first assembly for low-speed translational displacement and dropping of a control rod comprises a leak-tight casing tube which communicates with the reactor vessel, a screw and nut system coupled to the control rod by means of the nut and a screw -retaining device which is capable of releasing the screw. In the event of accidental failure of the casing tube, upward displacement of the control rod and of the broken portion of the casing tube is limited by a second assembly comprising means for rotational locking of the screw drive rotor and calibrated elastic means for producing action on the rotor in opposition to the vertical thrust produced by failure of the casing tube.

  14. Evaluation of LLNL's Nuclear Accident Dosimeters at the CALIBAN Reactor September 2010

    SciTech Connect (OSTI)

    Hickman, D P; Wysong, A R; Heinrichs, D P; Wong, C T; Merritt, M J; Topper, J D; Gressmann, F A; Madden, D J

    2011-06-21T23:59:59.000Z

    The Lawrence Livermore National Laboratory uses neutron activation elements in a Panasonic TLD holder as a personnel nuclear accident dosimeter (PNAD). The LLNL PNAD has periodically been tested using a Cf-252 neutron source, however until 2009, it was more than 25 years since the PNAD has been tested against a source of neutrons that arise from a reactor generated neutron spectrum that simulates a criticality. In October 2009, LLNL participated in an intercomparison of nuclear accident dosimeters at the CEA Valduc Silene reactor (Hickman, et.al. 2010). In September 2010, LLNL participated in a second intercomparison of nuclear accident dosimeters at CEA Valduc. The reactor generated neutron irradiations for the 2010 exercise were performed at the Caliban reactor. The Caliban results are described in this report. The procedure for measuring the nuclear accident dosimeters in the event of an accident has a solid foundation based on many experimental results and comparisons. The entire process, from receiving the activated NADs to collecting and storing them after counting was executed successfully in a field based operation. Under normal conditions at LLNL, detectors are ready and available 24/7 to perform the necessary measurement of nuclear accident components. Likewise LLNL maintains processing laboratories that are separated from the areas where measurements occur, but contained within the same facility for easy movement from processing area to measurement area. In the event of a loss of LLNL permanent facilities, the Caliban and previous Silene exercises have demonstrated that LLNL can establish field operations that will very good nuclear accident dosimetry results. There are still several aspects of LLNL's nuclear accident dosimetry program that have not been tested or confirmed. For instance, LLNL's method for using of biological samples (blood and hair) has not been verified since the method was first developed in the 1980's. Because LLNL and the other DOE participants were limited in what they were allowed to do at the Caliban and Silene exercises and testing of various elements of the nuclear accident dosimetry programs cannot always be performed as guests at other sites, it has become evident that DOE needs its own capability to test nuclear accident dosimeters. Angular dependence determination and correction factors for NADs desperately need testing as well as more evaluation regarding the correct determination of gamma doses. It will be critical to properly design any testing facility so that the necessary experiments can be performed by DOE laboratories as well as guest laboratories. Alternate methods of dose assessment such as using various metals commonly found in pockets and clothing have yet to be evaluated. The DOE is planning to utilize the Godiva or Flattop reactor for testing nuclear accident dosimeters. LLNL has been assigned the primary operational authority for such testing. Proper testing of nuclear accident dosimeters will require highly specific characterization of the pulse fields. Just as important as the characterization of the pulsed fields will be the design of facilities used to process the NADs. Appropriate facilities will be needed to allow for early access to dosimeters to test and develop quick sorting techniques. These facilities will need appropriate laboratory preparation space and an area for measurements. Finally, such a facility will allow greater numbers of LLNL and DOE laboratory personnel to train on the processing and interpretation of nuclear accident dosimeters and results. Until this facility is fully operational for test purposes, DOE laboratories may need to continue periodic testing as guests of other reactor facilities such as Silene and Caliban.

  15. Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

    SciTech Connect (OSTI)

    Not Available

    1992-07-01T23:59:59.000Z

    Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility.

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

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2003-01-01T23:59:59.000Z

    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...

  17. Standard Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E 706 (IH)

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2010-01-01T23:59:59.000Z

    1.1 This guide discusses test procedures that can be used in conjunction with, but not as alternatives to, those required by Practices E185 and E2215 for the surveillance of nuclear reactor vessels. The supplemental mechanical property tests outlined permit the acquisition of additional information on radiation-induced changes in fracture toughness, notch ductility, and yield strength properties of the reactor vessel steels. 1.2 This guide provides recommendations for the preparation of test specimens for irradiation, and identifies special precautions and requirements for reactor surveillance operations and postirradiation test planning. Guidance on data reduction and computational procedures is also given. Reference is made to other ASTM test methods for the physical conduct of specimen tests and for raw data acquisition.

  18. Estimates of the financial consequences of nuclear-power-reactor accidents

    SciTech Connect (OSTI)

    Strip, D.R.

    1982-09-01T23:59:59.000Z

    This report develops preliminary techniques for estimating the financial consequences of potential nuclear power reactor accidents. Offsite cost estimates are based on CRAC2 calculations. Costs are assigned to health effects as well as property damage. Onsite costs are estimated for worker health effects, replacement power, and cleanup costs. Several classes of costs are not included, such as indirect costs, socio-economic costs, and health care costs. Present value discounting is explained and then used to calculate the life cycle cost of the risks of potential reactor accidents. Results of the financial consequence estimates for 156 reactor-site combinations are summarized, and detailed estimates are provided in an appendix. The results indicate that, in general, onsite costs dominate the consequences of potential accidents.

  19. Nuclear reactor control rod cluster for enthalpy rise compensation

    SciTech Connect (OSTI)

    Freeman, T.R.

    1987-07-07T23:59:59.000Z

    This patent describes a control arrangement for use with a nuclear fuel assembly having guide thimbles to provide enthalpy rise compensation, comprising: (a) a spider assembly; and (b) control rodlets containing neutron absorber material, supported at their upper ends from the spider assembly in a pattern which matches that of the guide thimbles and extending generally parallel to one another downwardly for insertion into the guide thimbles of the fuel assembly; (c) the control rodlets arranged in a cluster having a higher axial neutron absorption capability at an upper portion than at a lower portion; each control rodlet varying in axial neutron absorption capability from every other rodlet.

  20. Characterization of nuclear reactor containment penetrations. Preliminary report

    SciTech Connect (OSTI)

    Bump, T.R.; Seidensticker, R.W.; Shackelford, M.A.; Gambhir, V.K.; McLennan, G.L.

    1984-06-01T23:59:59.000Z

    This report summarizes the survey work conducted by Argonne National Laboratory on the design and details of major penetrations in 22 nuclear power plants. The survey includes all containment types and materials in current use. It also includes details of all types of penetrations (except for electrical penetration assemblies and valves) and the seals and gaskets used in them. The report provides a test matrix for testing major penetrations and for testing seals and gaskets in order to evaluate their leakage potential under severe accident conditions.

  1. Spark Plasma Sintering of Fuel Cermets for Nuclear Reactor Applications

    SciTech Connect (OSTI)

    Yang Zhong; Robert C. O'Brien; Steven D. Howe; Nathan D. Jerred; Kristopher Schwinn; Laura Sudderth; Joshua Hundley

    2011-11-01T23:59:59.000Z

    The feasibility of the fabrication of tungsten based nuclear fuel cermets via Spark Plasma Sintering (SPS) is investigated in this work. CeO2 is used to simulate fuel loadings of UO2 or Mixed-Oxide (MOX) fuels within tungsten-based cermets due to the similar properties of these materials. This study shows that after a short time sintering, greater than 90 % density can be achieved, which is suitable to possess good strength as well as the ability to contain fission products. The mechanical properties and the densities of the samples are also investigated as functions of the applied pressures during the sintering.

  2. 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.

    2014-06-10T23:59:59.000Z

    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.

  3. 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

    2013-11-05T23:59:59.000Z

    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.

  4. Magnetic Field Generation in Planets and Satellites by Natural Nuclear Fission Reactors

    E-Print Network [OSTI]

    J. Marvin Herndon

    2007-09-27T23:59:59.000Z

    One of the most fundamental problems in physics has been to understand the nature of the mechanism that generates the geomagnetic field and the magnetic fields of other planets and satellites. For decades, the dynamo mechanism, thought to be responsible for generating the geomagnetic field and other planetary magnetic fields, has been ascribed to convection in each planet's iron-alloy core. Recently, I described the problems inherent in Earth-core convection and proposed instead that the geomagnetic field is produced by a dynamo mechanism involving convection, not in the fluid core, but in the electrically conductive, fluid, fission-product sub-shell of a natural nuclear fission reactor at the center of the Earth, called the georeactor. Here I set forth in detail the commonality in the Solar System of the matter like that of the inside of the Earth, which is my basis for generalizing the concept of planetary magnetic field generation by natural planetocentric nuclear fission reactors.

  5. Intact and Degraded Component Criticality Calculations of N Reactors Spent Nuclear Fuel

    SciTech Connect (OSTI)

    L. Angers

    2001-01-31T23:59:59.000Z

    The objective of this calculation is to perform intact and degraded mode criticality evaluations of the Department of Energy's (DOE) N Reactor Spent Nuclear Fuel codisposed in a 2-Defense High-Level Waste (2-DHLW)/2-Multi-Canister Overpack (MCO) Waste Package (WP) and emplaced in a monitored geologic repository (MGR) (see Attachment I). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (k{sub eff}) for both intact and degraded mode internal configurations of the codisposal waste package. This calculation will support the analysis that will be performed to demonstrate the technical viability for disposing of U-metal (N Reactor) spent nuclear fuel in the potential MGR.

  6. The Use of experiments on a single fuel element to determine the nuclear parameters of reactor lattices

    E-Print Network [OSTI]

    Pilat, E. E., 1937-

    1967-01-01T23:59:59.000Z

    The nuclear parameters of a reactor lattice may be determined by critical experiments on that lattice, by theoretical calculations in which only cross sections are used as input, or by methods which combine theory and ...

  7. CFD Analysis of Core Bypass Flow and Crossflow in the Prismatic Very High Temperature Gas-cooled Nuclear Reactor 

    E-Print Network [OSTI]

    Wang, Huhu 1985-

    2012-12-13T23:59:59.000Z

    Very High Temperature Rector (VHTR) had been designated as one of those promising reactors for the Next Generation (IV) Nuclear Plant (NGNP). For a prismatic core VHTR, one of the most crucial design considerations is the bypass flow and crossflow...

  8. TITAN : an advanced three dimensional coupled neutronicthermal-hydraulics code for light water nuclear reactor core analysis

    E-Print Network [OSTI]

    Griggs, D. P.

    1984-01-01T23:59:59.000Z

    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 ...

  9. Comments on: Small Modular Reactors

    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: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625govInstrumentstdmadapInactiveVisiting the TWPSuccessAlamosCharacterization2Climate,CobaltColdin679April

  10. Integral Validation of Minor Actinide Nuclear Data by using Samples Irradiated at Dounreay Prototype Fast Reactor

    SciTech Connect (OSTI)

    Tsujimoto, Kazufumi; Oigawa, Hiroyuki; Shinohara, Nobuo [Japan Atomic Energy Research Institute, Shirakata Shirane 2-4, Tokai, Ibaraki 319-1195 (Japan)

    2005-05-24T23:59:59.000Z

    The reliability of nuclear data for minor actinides was evaluated by using the results of the post-irradiation experiment for actinide samples irradiated at the Dounreay Prototype Fast Reactor. The burnup calculations with JENDL-3.3, ENDF/B-VI.8, and JEFF-3.0 were performed. From the comparison between the experimental data and the calculational results, in general, the reliability of nuclear data for the minor actinides are at an adequate level for the conceptual design study of transmutation systems. It is, however, found that improvement of the accuracy is necessary for some nuclides, such as 238Pu, 242Pu, and 241Am.

  11. Hydrogen loops in existing reactors for testing fuel elements for nuclear propulsion

    SciTech Connect (OSTI)

    Olsen, C.S.; Welland, H.; Abraschoff, J. (Idaho National Engineering Laboratory, EG G Idaho Inc., P.O. Box 1625, Idaho Falls, Idaho 83415 (United States)); Thoms, K. (Oak Ridge National Laboratory, P.O. Box, Oak Ridge, Tennessee 37831-8087 (United States))

    1993-01-15T23:59:59.000Z

    The Space Exploration Initiative (SEI) has revitalized interest in adapting nuclear energy for power and propulsion. Prior to the selection of a nuclear thermal propulsion (NTP) system, extensive testing of the various proposed concepts will be required. In today's environmental, safety and health culture, full size rocket engine tests as were done under the Rover/NERVA program will be extremely difficult and expensive to perform and meet NASA's schedules. A different test strategy uses a hydrogen loop in an existing reactor to test a wide variety of single elements or clusters of elements for fuel qualification. This approach is expected to reduce operating and capital costs and expedite the testing schedule. This paper examines the potential of performing subscale tests in a hydrogen loop in an existing reactor such as the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory or the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory. The HFIR is expected to achieve power densities comparable to those achieved in ATR because of the 85 MWt power level and the high thermal and fast flux levels. The available length and diameter of the test region of FHIR are 60 cm and 10 cm whereas the available length and diameter of the test region of ATR are 120 cm and 12 cm respectively.

  12. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

    SciTech Connect (OSTI)

    Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E. [Sandia National Labs., Albuquerque, NM (United States)] Sandia National Labs., Albuquerque, NM (United States); Tills, J. [J. Tills and Associates, Inc., Sandia Park, NM (United States)] J. Tills and Associates, Inc., Sandia Park, NM (United States)

    1997-12-01T23:59:59.000Z

    The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.

  13. Confirmatory Survey Results for the Reactor Building Dome Upper Surfaces, Rancho Saco Nuclear Generating Station

    SciTech Connect (OSTI)

    Wade C. Adams

    2006-10-25T23:59:59.000Z

    Results from a confirmatory survey of the upper structural surfaces of the Reactor Building Dome at the Rancho Seco Nuclear Generating Station (RSNGS) performed by the Oak Ridge Institute for Science and Education for the NRC. Also includes results of interlaboratory comparison analyses on several archived soil samples that would be provided by RSNGS personnel. The confirmatory surveys were performed on June 7 and 8, 2006.

  14. Descriptions of selected accidents that have occurred at nuclear reactor facilities

    SciTech Connect (OSTI)

    Bertini, H.W.

    1980-04-01T23:59:59.000Z

    This report was prepared at the request of the President's Commission on the Accident at Three Mile Island to provide the members of the Commission with some insight into the nature and significance of accidents that have occurred at nuclear reactor facilities in the past. Toward that end, this report presents a brief description of 44 accidents which have occurred throughout the world and which meet at least one of the severity criteria that were established.

  15. Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

    DOE Patents [OSTI]

    Crawford, Douglas C. (Idaho Falls, ID); Porter, Douglas L. (Idaho Falls, ID); Hayes, Steven L. (Idaho Falls, ID); Hill, Robert N. (Bolingbrook, IL)

    1999-01-01T23:59:59.000Z

    A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr--Hf alloy or an alloy of Pu--Zr--Hf or a combination of both.

  16. Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

    SciTech Connect (OSTI)

    Crawford, Douglas C.; Porter, Douglas L.; Hayes, Steven L.; Hill, Robert N.

    1997-12-01T23:59:59.000Z

    A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr-Hf alloy or an alloy of Pu-Zr-Hf or a combination of both.

  17. Optimal control theory for xenon spatial oscillations in load follow of a nuclear reactor

    SciTech Connect (OSTI)

    Cho, N.Z.

    1980-01-01T23:59:59.000Z

    A simple core control model has been developed for the control of xenon spatial oscillations in load-following operations of a current design commercial nuclear pressurized water reactor. The model has been formulated as a linear - quadratic - tracking problem in the context of moderm optimal control theory and the resulting two-point boundary value problem (TPBVP) has been solved directly by the techniques of initial value methods.

  18. Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

    DOE Patents [OSTI]

    Crawford, D.C.; Porter, D.L.; Hayes, S.L.; Hill, R.N.

    1999-03-23T23:59:59.000Z

    A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr--Hf alloy or an alloy of Pu--Zr--Hf or a combination of both. 7 figs.

  19. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Main report

    SciTech Connect (OSTI)

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01T23:59:59.000Z

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). The study results are presented in two volumes. Volume 1 (Main Report) contains the results in summary form.

  20. Prospects for and problems of using light-water supercritical-pressure coolant in nuclear reactors in order to increase the efficiency of the nuclear fuel cycle

    SciTech Connect (OSTI)

    Alekseev, P. N.; Semchenkov, Yu. M.; Sedov, A. A., E-mail: sedov@dhtp.kial.ru; Subbotin, S. A.; Chibinyaev, A. V. [Russian Research Centre Kurchatov Institute (Russian Federation)

    2011-12-15T23:59:59.000Z

    Trends in the development of the power sector of the Russian and world power industries both at present time and in the near future are analyzed. Trends in the rise of prices for reserves of fossil and nuclear fuels used for electricity production are compared. An analysis of the competitiveness of electricity production at nuclear power plants as compared to the competitiveness of electricity produced at coal-fired and natural-gas-fired thermal power plants is performed. The efficiency of the open nuclear fuel cycle and various versions of the closed nuclear fuel cycle is discussed. The requirements on light-water reactors under the scenario of dynamic development of the nuclear power industry in Russia are determined. Results of analyzing the efficiency of fuel utilization for various versions of vessel-type light-water reactors with supercritical coolant are given. Advantages and problems of reactors with supercritical-pressure water are listed.